Inquiry into the radiological consequences of power uprates at light-water rectors worldwide

Inquiry into the radiological consequences of power

uprates at light-water rectors worldwide

Tea Bilic Zabric, Bojan Tomic, Klas Lundgren and Mats Sjöberg

SSI Rapport 2007:07

Rapport från Statens strålskyddsinstituttillgänglig i sin helhet via www.ssi.se

Ultraviolet, solar and optical radiationUltraviolet radiation from the sun and solariums can result in both long-term and short-term effects. Other types of optical radiation, primarily from lasers, can also be hazardous. SSI provides guidance and information.

SolariumsThe risk of tanning in a solarium are probably the same as tanning in natural sunlight. Therefore SSI’s regulations also provide advice for people tanning in solariums.

RadonThe largest contribution to the total radiation dose to the Swedish population comes from indoor air. SSI works with risk assessments, measurement techniques and advises other authorities.

Health careThe second largest contribution to the total radiation dose to the Swedish population comes from health care. SSI is working to reduce the radiation dose to employees and patients through its regulations and its inspection activities.

Radiation in industry and researchAccording to the Radiation Protection Act, a licence is required to conduct activities involving ionising radiation. SSI promulgates regulations and checks compliance with these regulations, conducts inspections and investigations and can stop hazardous activities.

Nuclear powerSSI requires that nuclear power plants should have adequate radiation protection for the generalpublic, employees and the environment. SSI also checks compliance with these requirements on a continuous basis.

WasteSSI works to ensure that all radioactive waste is managed in a manner that is safe from the standpoint of radiation protection.

Mobile telephonyMobile telephones and base stations emit electromagnetic fields. SSI is monitoring developments and research in mobile telephony and associated health risks.

TransportSSI is involved in work in Sweden and abroad to ensure the safe transportation of radioactive substances used in the health care sector, industrial radiation sources and spent nuclear fuel.

Environment“A safe radiation environment” is one of the 15 environmental quality objectives that the Swedish parliament has decided must be met in order to achieve an ecologically sustainable development in society. SSI is responsible for ensuring that this objective is reached.

BiofuelBiofuel from trees, which contains, for example from the Chernobyl accident, is an issue where SSI is currently conducting research and formulating regulations.

Cosmic radiationAirline flight crews can be exposed to high levels of cosmic radiation. SSI participates in joint international projects to identify the occupational exposure within this job category.

Electromagnetic fieldsSSI is working on the risks associated with electromagnetic fields and adopts countermea-sures when risks are identified.

Emergency preparednessSSI maintains a round-the-clock emergency response organisation to protect people and the environment from the consequences of nuclear accidents and other radiation-related accidents.

SSI Education is charged with providing a wide range of education in the field of radiation protection. Its courses are financed by students' fees.

SSI's Activity Symbols

SSI rapport: 2007:07

maj 2007

ISSn 0282-4434

The conclusions and viewpoints presented in the report are those of the authors and do not necessarily coincide with those of the SSI.

Författarna svarar själva för innehållet i rapporten.

edItorS / redaktörer : Tea Bilic Zabric1), Bojan Tomic1), Klas Lundgren2) and Mats Sjöberg3) 1)Enconet Consulting, 2)ALARA Enginering and 3)ES-konsult

tItle / tItel: Inquiry into the radiological consequences of power uprates at light-water rectors worldwide. / Utredning av radiologiska konsekvenser i sam-band med effekthöjningar i lättvattenreaktorer världen över.

department / avdelnIng: Department of Occupational and Medical Exposures / Avdelning för för personal- och patientstrålskydd.

Summary: In Sweden, most of the nuclear power plants are planning power uprates within the next few years. The Department of Occupational and Medical Exposures at the Swedish Radiation Protection Agency, SSI, has initiated a research project to investigate the radiological implications of power uprates on light-water reactors throughout the world.

The project was divided into three tasks:

1. A compilation of power uprates of light-water reactors worldwide. The compi-lation contains a technical description in brief of how the power uprates were carried out.

2. An analysis of the radiological consequences at four selected Nuclear Power Plants, which was the main objective of the inquiry. Affects on the radiolo-gical and chemical situation due to the changed situation were discussed.

3. Review of technical and organisational factors to be considered in uprate projects to keep exposures ALARA.

The project was carried out, starting with the collecting of information on the imple-mented and planned uprates on reactors internationally. The information was catalogued in accordance with criteria focusing on radiological impact. A detailed analysis followed of four plants selected for uprates chosen according to established criteria, in line with the project requirements. The selected plants were Olkiluoto 1 and 2, Cofrentes, Asco and Tihange. The plants were selected with design and operation conditions close to the Swedish plants. All information was compiled to identify good and bad practices that are impacting on the occupa-tional exposure. Important factors were discussed concerning BWRs and PWRs which affect radiation levels and occupational exposures in general, and especially at power uprates.

Conclusions related to each task are in detail presented in a particular chapter of the report. Taking into account the whole project and its main objective the following conclu-sions are considered to be emphasized

Optimisation of the work processes to limit the duration of the time spent in the con-trolled areas is especially important. Leadership, composition and organization of the large demanding tasks are critical for successful implementation of power uprate and keeping received doses at a minimum. Good planning and preparation, which reflects experience from similar projects elsewhere, adherence to procedures and supervision from plant personnel as well as consequential application of ALARA principles and good practices are important factors.

It has not been found a direct relationship between the uprates and the occupational expo-sures. The occupational doses on some plants seem to be higher after the uprate, while on others seem to be lower. However the general trend in light-water reactors worldwide is gradually reduced occupational exposures.

There is no obvious correlation of the power uprate and fuel failures. However, per-form-ance of fuel for PWRs and BWRs went in opposing directions, improving for PWRs and deteriorating for BWRs.

For BWRs investment in the condensate cleanup efficiency results in favourable water chemistry conditions that can be maintained, or even improved, after the power uprate. The higher steam velocity after a power up-rate can increase the radiation levels around main steam lines and other turbine components due to a consider-able increase in steam moisture content. This problem can be overcome with a recent design and installation of new steam dryers in the reactor pressure vessel to reduce steam moisture.

Issues of relevance for PWRs include: Increase in the rate of production of H-3 due to higher boron concen-tration and power level, especially for longer fuel cycles; Control of pH and Lithium as an essential means of controlling the corrosion level and thus radiation levels. Fuel related corrosion problems are shown to be less visible with good pH control and shorter fuel cycles.

SammanfattnIng: I Sverige pågår eller planeras effekthöjningsprojekt i de flesta av kärnkraftverken. Sta-tens strålskyddsinstituts avdelning för personal- och patientstrålskydd har initierat ett forskningsprojekt för att samla in information om effekthöjningar i lättvattenreaktorer världen över med fokus på de radiologiska konsekvenserna som effekthöjningen innebär.

Projektet delades in i tre delprojekt:

1. Insamling av fakta om reaktorer som höjt eller står i begrepp att höja effekten med en kortfattad teknisk beskrivning om effekthöjningen.

2. Analys av de radiologiska konsekvenserna på fyra utvalda kärnkraftverk, vilket var det egentliga huvudsyftet med projektet. Påverkan på den radiologiska och kemiska situationen pga. den änd-rade situationen diskuterades.

3. Genomgång av tekniska och organisatoriska faktorer som bör beaktas vid en ef-fekthöjning för att hålla stråldoser så låga som rimligen är möjligt (ALARA).

Projektet genomfördes med att initialt samla in internationell information om genomförda och planerade effekthöjningar. Informationen katalogiserades i enlighet med kriterier som fokuserade på radiologisk påver-kan. En detaljerad analys genomfördes av fyra utvalda kärnkraftverk valda enligt fastställda kriterier som uppfyllde projektets anspråk. De valda kraftverken var Olkiluoto 1 and 2, Cofrentes, Asco and Tihange. Des-sa kraftverk valdes med hänsyn till att konstruktion och driftbetingelser ligger nära de svenska kraftverken. Information samlades in för att identifiera “good and bad practices” som påverkar perso-nalens stråldoser. Viktiga faktorer som diskuterades berör strålnivåer och personalens stråldoser generellt och vid effekthöj-ningar i synnerhet.

Varje delprojekts slutsatser är presenterade i delprojektets kapitlet. Vi vill dock med hän-syn till projektet och dess mål framhålla följande slutsatser:

Optimering av arbetsprocesser för att begränsa tiden på kontrollerat område är av stor vikt. Ledarskap, sam-mansättning och organisation av stora utmanande projekt är kritiska faktorer för att genomföra en effekt-höjning med lyckat resultat. God planering och förbe-redelse som tar hänsyn till liknande projekt vid andra anläggningar, användning av in-struktioner och att kraftverkets egen personal utför en noggrann övervakning av arbeten samt införande av ALARA principer och användandet av goda förebilder är viktiga fakto-rer.

Någon direkt relation mellan effekthöjning och personalens stråldoser har inte kunnat påvisas. Personalens stråldoser på några kraftverk verkar öka medan på andra sjunker doserna. En generell trend i lättvattenvat-tenreaktorer världen över är gradvis minskade doser till personalen.

Det har inte kunnat påvisas någon tydlig korrelation mellan effekthöjning och bränsleska-dor. Emellertid har bränsleskadeutvecklingen för PWR och BWR gått motsatta vägar. För PWR har den förbättrats men för BWR har den försämrats.

För BWR har det visat sig att investeringar i att effektivisera kondensatreningssystemet har resulterat i lika bra eller t.o.m. bättre kemiförhållande. Den högre ånghastigheten efter en effekthöjning kan resultera i högre strålnivåer runt ångledningarna och turbinkompo-nenter pga. ökad fukthalt i ångan. Detta problem kan åt-gärdas genom installation av en modern konstruktion av ångseparator i reaktortanken.

I tryckvattenreaktorer ökar H-3 produktionen genom högre effektnivå och ökad borkon-centration, speciellt med långa driftcykler. Kontroll av pH- och litiumnivå är väsentliga hjälpmedel för kontroll av korrosion och därigenom strålnivån. Problem med bränslet kan undvikas genom god kontroll av pH och begränsning av bränslecykellängden.

1

Table of contents

List of abbreviations ............................................................................................................2

1. Introduction .................................................................................................................5

2. Project description .......................................................................................................6

2.1 Project overview..................................................................................................6

2.2 Working method..................................................................................................6

2.3 Participants ..........................................................................................................7

3. Compilation of power uprates ....................................................................................8

3.1 Introduction ........................................................................................................8

3.2 Data Collection and Sources...............................................................................8

3.3 Content of the data base....................................................................................11

3.4 Conclusions ......................................................................................................15

4. Analysis of the selected plants..................................................................................21

4.1 Olkiluoto 1 and 2 ..............................................................................................21

4.1.1 Introduction ..............................................................................................21

4.1.2 OL1/2 power uprate..................................................................................21

4.2 Cofrentes...........................................................................................................53

4.2.1 Introduction ..............................................................................................53

4.2.2 CNC power uprate ....................................................................................53

4.3 Asco and Tihange .............................................................................................78

4.3.1 Introduction ..............................................................................................78

4.3.2 Asco NPP power uprate............................................................................81

4.3.3 Tihange NPP power uprate.......................................................................89

4.3.4 Comparison of two PWR uprates .............................................................96

5. Reconstruction experience......................................................................................103

5.1 Technical factors ............................................................................................103

5.1.1 Introduction ............................................................................................103

5.1.2 BWR uprates...........................................................................................103

5.1.3 PWR uprates ............................................................................................122

5.2 Organisational factors.....................................................................................134

5.2.1 Reducing the exposures ..........................................................................134

5.2.2 Implementation of power uprate.............................................................136

5.2.3 Lesson learned ........................................................................................138

6. Conclusions .............................................................................................................140

7. References ..............................................................................................................147

Appendix 1: Data base with NPP uprates........................................................................149

2

List of abbreviations

AB Auxiliary building

ABB ASEA merged with BBC

ACEC Alstom ACEC Energie

ACLF Grouping of ACEC/C-L//Framatome/WNE

AEE Atomenergoexport

AFW Auxiliary Feedwater

ALARA As Low As Reasonably Achievable

AO Axial Offset

AOA Axial Offset Anomaly

ASEA

Allmänna Svenska Elektriska Aktiebolaget (General Swedish Electrical

Limited Company)

BBC Brown Boverie et Cie

BOC Beginning Of Cycle

BRAC BWR Radiation Assessment and Control

BWR Boiling Water Reactor

°C Degrees Centigrade

CCU Condensate Clean-Up

CE Combustion Engineering

CILC Crud Induced Local Corrosion

C-L Creusot Loire

CMI Cockerill Mechanical Industry

CNC Confrentes nuclear power plant

CRD Control Rod Drives

CVCS Chemical Volume Control System

DB Deep Bed

DG Diesel Generator

DH Dissolved Hydrogen

DO Dissolved Oxygen

dP Delta Pressure

DR Dose Rate

E Extended power

EPU Extended Power Uprate

DZO Depleted Zinc Oxide

EBA Enriched Boron Acid

EC European Commission

ECP Electrochemical (or corrosion) Potential

EFPH Effective Full Power Hour

EOC End Of Cycle

EPRI Electric Power Research Institute (www.epri.com)

°F Degrees Fahrenheit

F + DB Filter + Deep Bed

FD Filter Demineralizer

FPHD Forward Pumped Heater Drains

3

FRAM Framatome

FRAMACECO Framatome-ACEC-CO

FW Feedwater

GBq Giga Becquerel

GE General Electric

Gpm Gallons per minute

HP High Pressure

HWC

Hydrogen Water Chemistry in BWRs with injection of hydrogen in order

to reduce the risk of environmental assisted cracking

HWC-M Moderate HWC

IAEA International Atomic Energy Agency

ICRP International Commission on Radiological Protection

IGSCC Intergranular Stress Corrosion Cracking

ISOE Information System on Occupational Exposure

KWU Kraftwerk Union

LEFM Ultrasonic feedwater flow measuring system

LP Low Pressure

MP Measuring Position for dose rate

MSL Main Steam Line

MSLR Main Steam Line Radiation

MU Measurument Uncertainty

MUR Measurument Uncertainty Recapture

MWt MW thermal power

NM Noble Metal

NMCA Noble Metal Chemical Addition

NPP Nuclear Power Plant

NRC Nuclear Regulatory Commission

NSSS Nuclear steam supply system

NWC Normal Water Chemistry in BWRs without injection of hydrogen

O&M Operation and Maintenace

OECD Organisation for Economic Cooperation and Development

OL1 Olkiluoto-1

OL2 Olkiluoto-2

OLNC On Line NobleChem

PCI Pellet Cladding Interaction

PCMI Pellet Cladding Mechanical Interaction

PCS Power Conversion System

pH300 pH at 300 �C

PI Performance Indicators

PLR Primary Loop Recirculation (i.e. recirculation lines)

PS Pressure Suppression

PWR Pressurized Water Reactor

R2 Ringhals 2

R3 Ringhals 3

R4 Ringhals 4

RB Reactor Building

4

RCS Reactor coolant system

RFO Refuelling Outage

RHR Residual Heat Removal

Rpm Rounds per minute

RPV Reactor Pressure Vessel

RTD Reactor Temperature

RW Reactor Water

RWCU Reactor Water Clean-Up

Rx Reactor

S Stretch power

SG Steam Generator

SGR Steam Generator Replacement

SHE Standard Hydrogen Electrode

SPF Spent Fuel Pit

SPU Stretch Power Uprate

SS Stainless Steel

SSI Swedish Radiation Protection Authority

STP Standard Temperature and Pressure

T Ave Average temperature

T½ Half-life for radionuclide

TMI Three Mile Island

TVO Teollisuuden Voima Oy

WANO World Association of Nuclear Operators (www.wano.org.uk)

WEC Westinghouse

WNE Westinghouse Nuclear

XO Extra Outage

5

1. Introduction

Most nuclear power plants in Sweden are planning power uprates within the next few

years. Permission to increase power is given by the Swedish Government. The Swedish

Radiation Protection Authority, SSI, is one of the bodies to which the application for

power uprates is referred for consideration. The Department of Occupational and Medical

Exposures at SSI has initiated an inquiry to consider the radiological implications of

thermal power uprates on light-water reactors throughout the world. The information

gained from the research will be firstly used as a reference and background information

source and then toreview the the applications for power uprates and in the assessment of

the after-effects of these uprates.

Available information shows that a relatively high percentage of all operating NPPs in the

world have implemented, or are considering, some form of power increase (power

uprate). Such uprates vary significantly. Small uprates of a few percent of the plant’s

power may be achieved by modification of the power conversion system and/or adjust-

ments to control systems. Conversley large uprates, sometimes in excess of 30% nominal

power, may be undertaken which require substantial changes on the reactor side, includ-

ing fuel, operating regime and limits, etc.

The majority of power uprates are in the middle range (between 5 and 10 % of nominal

power) and typically involve changes to both reactor and power conversion system

(PCS). However, all power uprates require either major or minor modification to operat-

ing practices and conditions. The radiological doses to personnel are related to these up-

grades, both during normal operation and during outages, whilst also being sensitive to

differing materials and operating regimes. Integral doses could often be found from

WANO indicators and other sources of information. However, these had not been sys-

tematically analysed to determin which specific features of the uprates were influencing

radiological doses.

6

2. Project description

2.1 Project overview

The aim of the inquiry was to investigate what specific conditions and practices affect the

operational doses received when reactor power is uprated. Identification of these factors

on a worldwide basis should then allow power uprates to be planned in way that provides

better exposure optimisation.

The inquiry was divided into three tasks:

1. A compilation of power uprates of light-water reactors worldwide. The compila-

tion contains a technical description in brief of how the power uprates were car-

ried out.

2. The main emphasis of the inquiry was an analysis of the radiological conse-

quences at four selected Nuclear Power Plants. Affects on the radiological situa-

tion due to the changed situation was discussed by checking areas of special in-

terest, such as

− degradation of material resulting in more repair work,

− verification of safety and security resulting in more testing and

− work performed in controlled areas in relation to the uprate.

3. Experience from the reconstruction period with bearing on the radiation protec-

tion of workers

This report is a compilation of all three tasks. Each task has its own chapter and for task 2

the analysis of the selected plants, are shown in three different subchapters. Task 3 is

divided into two subchapters where the technical factors to control radiation fields are

discussed in one and the organisational issues in the second.

2.2 Working method

This inquiry was implemented with three specific elements (tasks), starting with the col-

lecting of information on the implemented and planned uprates on PWR and BWR reac-

tors internationally. The information was catalogued in accordance with criteria focusing

on radiological impact. A detailed analysis of plants selected for uprates, were chosen

according to established criteria, in line with the project requirements. For BWR were

two plants selected, one with 12% power increase and another with 25%. For the PWRs

two uprates in the range of 10% power increase were selected. The plants were selected

with design and operation conditions close to the Swedish plants. The aim was a detailed

analysis of causal relations between uprates and the radiological content, thus the project

was organized into three tasks that were implemented one after another.

Data collection for the detailed analysis was carried out through personal contacts. Data

was specified which was specific for the particular type of reactor and sent to the contact

7

persons. There was a good response for the data sought for the BWR’s. For the PWRs

detailed radiological data was received but less technical and chemistry. The PWR Task 2

report is therefore, not the detailed analysis we hoped to make. Though a detailed analysis

on radiological data has been performed.

2.3 Participants

Three experienced companies carried out the inquiry. The experts involved in the project

were:

Tea Bilic Zabric and Bojan Tomic from Enconet Consulting, Vienna

Klas Lundgren from ALARA Enginnering, Skultuna

Mats Sjöberg ES-konsult, Solna

8

3. Compilation of power uprates

3.1 Introduction

Quite a few NPPs have implemented, or are considering increasing the power level on

which they operate (power uprate). Those power uprates vary significantly, from small

ones of a few percentages to large uprates above 30% of the nominal power. The altera-

tions to the plants particularly those with larger power uprates, require changes and modi-

fications to the plant, in both hardware and operating arrangements (procedures, operating

regime, and operating window). Apart from different requirements (on power conversion

system) fuel remains an important (and also a limiting) issue during an uprate.

In addition to the safety impact of uprates that is normally verified in depth before a li-

censee for an uprate is issued the radiological aspects, the consequences of an uprate are

also of interest. Changes in operating regimes, but also changes and operation with new

hardware might have an impact on the operational doses and their distribution. Increased

power level may also have certain impact on the effluents, especially on tritium.

The first task was to compile a database using worldwide sources. This database will be

accompanied by a discussion of the sources used and the initial conclusions that could be

drawn from the data.

3.2 Data Collection and Sources

The data collected within this task relates to the worldwide uprates performed (or

planned) and which shows the radiological consequences and compares the data before

and after the uprates. The data was collected from the literature sources; including a vari-

ety of databases, regulatory filings, analyses and other available information.

The following sources were used:

WANO Performance indicators

WANO maintains five programmes for information exchange, promoting mutual com-

munications and benchmarking. Two of them: ‘Exchange of Operating Experience’ and

‘Performance Indicators’ – a series of standardised parameters for the comparison of

power plants, were reviewed for data collection within this project. The data, readily

available from the WANO performance indicators database is more general and cannot be

used to determine doses in e.g. outages. The data is available from 1992. In some samples

for multi unit sites the doses in WANO indicators database are just a fraction (1/3 or 1/2)

of the plant’s total value. The WANO database was used for the initial review of occupa-

tional doses and to support some generic conclusions.

OECD ISOE data base

The ISOE Programme is the world’s largest occupational exposure database, established

by a network of radiation protection experts from operators and regulators. The ISOE

data serves as a point for the exchange of information and experience but it is also used to

support analysis. ISOE structure supports the collection of doses during outages with

specific doses for particular activities. The ISOE database distinguishes plant personnel

9

and subcontractors. ISOE database was established in 1992, but the data on collective

doses start in 1977.

The ISOE database was used to extract the information on occupational doses during

outages and during normal operation (annual doses).

Nuclear Engineering

The World Nuclear Industry Handbook is a reference guide to the nuclear power industry.

It is updated each year. Among other information the Handbook contains information on

power reactors, a country-by-country summary of reactors showing type, status, location,

main contractors and key dates; main data on each unit including technical detail on core,

vessel containment, fuel, coolant, moderator, control, fuelling, operating strategy, turbine

and more.

IAEA

The IAEA is a leading publisher in the nuclear field. It’s scientific and technical publica-

tions cover fifteen subject areas. They include the proceedings of major international

conferences, as well as international guides, codes, standards, reports, documents and

conventions. IAEA PRIS data base and publication was used for reviewing information

and proceedings from conferences.

NRC

The NRC's REIRS system provides the latest available information on radiation exposure

to the workforce at certain NRC licensed facilities. REIRS contains several data bases

that record the radiation exposure information. We used ‘Effluent Database for Nuclear

Power Plants’, which was developed to track annual aqueous and atmospheric effluent

release data and offsite doses calculated for each nuclear power plant in the United States.

The data is available from year 1998. The OECD developed the document ‘Thermal

Power Uprating in Europe’ which reflects the cooperation of many experts in Europe.. In

addition to the uprating data, also included was some ‘Plant data’ and general information

available from relevant countries, mostly based on the "IAEA PRIS" data.

EC

The Commission periodically publishes reports on releases to the environment of radioac-

tive substances in airborne and liquid effluents from Nuclear Power Stations and Nuclear

Fuel Reprocessing Sites in the European Union. These reports cover discharges from

Nuclear Power Stations of capacity greater than 50 MWe as well as from (former) Nu-

clear Fuel Reprocessing Sites.

Comments on the comparison of data across the sources of data

There is no systematic collection of data covering uprates and related activities, nor there

is any specific collection of radiological exposure information. Therefore, the information

of relevance for radiological impact of uprates was collected from a combination of

sources. While some of the sources were traditional ones, in others the data collection

started only recently. Because of that, a meaningful comparison in relation with earlier

uprates is not possible. The main source of data that was used for radiological releases

cover the years after 1995. Therefore, it was not possible to assess the effects of earlier

uprates. For the occupational exposures, the main source of data was the ISOE data base.

10

To assure the correctness and to be able to corroborate the ISOE data, a comparison with

the WANO data base entries for analysed plants was undertaken.

The table below compares ISOE and WANO data for the average recorded Occupational

doses during outages (for a period before and after an uprate) with the recorded Occupa-

tional dose for the outage during which an uprate was implemented. The assessment was

made for several plants that are comparable in their characteristics.

As can be seen from the data in the table, significant differences are visible in some cases

between ISOE and WANO data. Even after evaluating the reporting requirements, the

explanation for those differences could not be found.

To assure the consistency of any analysis within this project, a decision was taken to ex-

clusively use the ISOE data as a figure of merit for the occupations exposures during

normal operation and outages. The ISOE Programme is the world's largest collection of

information on occupational exposure. The ISOE data collection is structured in a way to

relate doses during outages with specific activities undertaken. Moreover the ISOE data

separate plant's personnel and subcontractors, thus allowing for a comparison of plants

that use external support differently.

Occupational dose in operating year Dose during uprate

(annual doses in year

uprate was implemented)

Before the uprate

(manSv)

After the uprate

(manSv)

manSv

Examples

ISOE data WANO

data

ISOE data WANO

data

ISOE data WANO

data

Plant1 1.202 1.14 0.879 0.83 3.30 1.49

Unit1/Plant2 1.282 0.99 0.677 0.78 3.22 2.03

Unit2/Plant2 1.432 0.99 0.986 0.78 1.41 2.03

Plant3 0.488 0.5 0.752 0.83 1.54 0.90

11

3.3 Content of the data base

Keeping in mind the overall objective of the project, the criteria for cataloguing the in-

formation to be collected was established. These reflected the knowledge of elements that

are impacting on the radiological doses, and that could be related to the uprate activities.

The printout of the Data base containing all information collected is provided in the Ap-

pendix 1.

The table below describes the fields that are included in the data base and discusses the

contents of each of those.

Field # Title Description/comments Main

Reference

0 Country Country where plant is located

1 Plant name

Plant name with unit indication

(if more uprates were performed, year of the

uprate follow the unit designator, i.e. Thiange

2/2001)

WANO PI

2.1 Vendor of NSSS Vendor (GE,WEC, FRAM, etc) WANO PI

2.2 Commercial data

of operation Month, Day, Year, WANO PI

3.1 Reactor type (PWR, BWR)

OECD,

Nuclear

Engineering

3.2 Initial power Original thermal power in MWt

OECD,

Nuclear

Engineering

4.1 Thermal power Uprated thermal power in MWt OECD, IAEA,

NRC

4.2 Year

implemented

Year when the uprate was implemented (in some

cases approved, data sometimes inconsistent) OECD, NRC

4.3

Uprate type &

total power

increase in the

percentage for

particular year

-MU -Measurement uncertainty (uprates are less

than 2 %)

-S -Stretch power (uprates typically up to 7)

-E - Extended power (uprates greater than the

stretch)

Example:

1985 S 4.1%

2001 E 19.4%

The first uprate was in 1985 (4.1%) and the

second in 2001 (15.4). The % always indicate the

total power increase compared with original

design (thus 19.4% in second)

OECD, NRC

4.4 Technical

solution

Technical solution for an uprate.

Example:

Whether the uprate was implemented by increase

of Rx T Avg (with the same FW mass flow) OR

T Avg remains same (but FW/MSL mass flow

and pressures were increased)

Data entered where available

OECD, NRC

4.5 Equipment Where available, list of main equipment

modified/affected OECD, NRC

12

Field # Title Description/comments Main

Reference

5.1 Fuel cycle Length of the fuel cycle - in the months Nuclear

Engineering

5.2

Average linear

fuel rating

(before uprate)

Fuel rating - in kW/m Nuclear

Engineering

5.3

Average linear

fuel rating (after

uprate)

Change in fuel rating – kW/m Nuclear

Engineering

5.4 Fuel type

Type of fuel used.

Sometimes different types used, depends on core

design. Equilibrium cycle fuel entered (when

available).

Nuclear

Engineering

6.1

Annual liquid

effluents (before

uprate)

Total Liquid release in GBq. (Data from EC

used, available from 1995. For US, data

available from 1998)

EC

6.2

Annual liquid

effluents (after

uprate)

As above EC

7.1

Annual gaseous

effluents (before

uprate)

Total Airborne releases in GBq EC

7.2

Annual gaseous

effluents (after

uprate)

As above EC

8.1

Annual

occupational

dose (before

uprate)

Average value of the total collective dose over

the three-year period before the uprate in manSv

(when three years not available, it is noted in

table)

ISOE, NRC

8.2

Annual

occupational

dose (after

uprate)

Average value of the total collective dose over

the three-year period after the uprate in manSv

(when three years not available, it is noted in

table)

ISOE, NRC

9.1

Occupational

dose during

outage (before

uprate)

Average value of the collective outage dose over

the three-year period before the uprate - in

manSv (when three years not available, it is

noted in table)

ISOE

9.2

Occupational

dose during

outage (after

uprate)

Average value of the collective outage dose over

the three-year period before the uprate - in

manSv (when three years not available, it is

noted in table)

ISOE

10

Occupational

dose during

uprate

Doses received during the uprate (if equipment

was changed) – in manSv.

Data not systematically available. In some cases

data cover annual occupational dose for the year

when the uprate was implemented.

Of no relevance for MU.

Of limited relevance for S, except when major

plant modifications implemented.

Sometimes involve SGR, if performed in parallel

ISOE

13

Classification of uprates

There are considerable economic benefits to uprates because they allow more value (en-

ergy) to be generated by the existing plant. Whilst the fuel costs may marginally rise the

remaining costs do not increase. This makes uprates highly attractive to the utilities. Nev-

ertheless, the complexity and significance of the safety and operational issues associated

with uprates make additional gains anything but easy. Comprehensive safety analysis and,

depending on the country, re-licensing by the regulator are important elements of every

uprate project. Changes in operating practices and methods of maintenance organisation

following uprates lead to radiological consequences both in normal operation and during

outages.

The aim of every uprate is to increase the electrical power output available from the main

generator. This can be achieved by modifying the power conversion system (e.g. turbine,

generator, associate equipment) and/or by increasing the reactor energy output.

The development of technology of, in particular, turbines in the last decade is such that

many plants increased the power by installing new turbines or parts of them, and achieved

the power increase of up to 3%. As this project is focused on radiological issues, the in-

crease of the generated energy through modifications on the PCS does not introduce any

effects of interest.

The second way to increase generation is to increase the power of the reactor. Typically

there are three distinctive categories of power increase (although in the first category,

MU, the reactor power is, physically, not increased), as follows:

Measurement Uncertainty (MU) Recapture Power Uprate: uprates of 1 to 2 percent

power, typically achieved using more precise techniques for measuring Feedwater flow

and/or performing analysis to reduce unnecessary conservatism

Stretch Power (S) Uprate: uprates of 5 to 7 percent power, typically achieved by chang-

ing instrumentation set points, re-analysis (to recover excessive margins) together with a

small number of major plant modifications

Extended Power (E) Uprate: uprates of up to 20 percent power, achieved by major

changes of core design and significant modifications to major plant equipment

The majority of power uprates implemented or planned are in the middle range of be-

tween 5 and 10 % nominal power. These typically involve changes to both the reactor and

the power conversion system. Some plants have performed different types of uprate on

two or more occasions.

Improved measurement and analysis techniques have allowed utilities to increase the

licensed power limits of existing plants as a cost-effective method of increasing power.

Currently, 76 PWR units (19 Europe and 57 USA) and 45 BWR units (11 Europe, 32

USA and 2 Mexico) have uprated thermal power. Of these, 24% were small uprates of up

to 2% increase, 49% were stretch and 27% were large power uprates.

A number of European and Asian utilities are planning to implement power uprates

within the next few years:

14

Dukovany 1-4 (PWR) is planing to increase power for 2.3% in the years 2005-2008

with installation of new turbine blades. It plans a 10% uprate at a later date.

Brokdorf (PWR) is planning 3.9% uprate (date not known)

Emsland (PWR) is planning 4.9% uprate (date not known)

Grafenrheinfeld (PWR) is planning 4.9% uprate (date not known)

Grohnde (PWR) is planning 4.5% uprate (date not known)

Gundremmingen B and C (BWR) are planning 6.8% uprate (date not known)

Isar-1 (BWR) is planning 7% uprate (date not known)

PAKS 1, 2, 4 (PWR) are planning 9.1% uprate in 2006

PAKS 3 (PWR) is planning 9.1% uprate in 2007

Kori 3 & 4 (PWR) are planning 5% uprate in 2006

YGN 1 & 2 (PWR) are planning 5% uprate in 2006

Higasidory (ABWR) is planning uprate (% and date not known)

Shika (BWR) is planning uprate (% and date not known)

Forsmark-1 (BWR) is planning 19.9% uprate in 2010



Oskarshamn-3 (BWR) is planning 29.8% uprate in 2008

Ringhals-1 (BWR) is planning 11.9% power uprate in not known

Ringhals-3 (PWR) is planning 7.8% uprate in year 2007 and 13.5% uprate in 2007

Ringhals -4 (PWR) is planning 13.5% uprate in 2011

The following table describing intended future power uprates in the USA is based on

information obtained from a survey of all licensees conducted in March 2006.

Fiscal Year Power Uprates

Expected

MUR SPU EPU MWt

2006 4 1 0 3 1470

2007 6 5 1 0 431

2008 0 0 0 0 0

2009 10 2 3 5 1792

2010 2 2 0 0 76

2011 1 1 0 0 26

15

3.4 Conclusions

Some conclusions are raised from the review of the data collected. Issues of interest are

discussed in section below.

Relevance of yearly occupational doses:

The data table provided the annual occupational doses for (usually) 3 years average be-

fore and after the uprate has been implemented. While this provides (some) insights re-

lated with occupation doses, the annual occupational doses are often driven by processes

and activities that have nothing to do with uprate, rather with specific repairs and inter-

ventions during outages and/or some specific operational issues (i.e. unusual leaks,

change of chemistry, material used, etc). In some case, unusual events might add to the

collective annual doses. While the averaging over a longer period (i.e. 3 years) remove

some of the impact of unusual events or specific repairs, it does not remove it completely.

Therefore, it is difficult to make any global conclusions and relate the uprate with any of

the annual occupation doses as documented in the data sources used.

Cycle length

In the last decade, many plants decided to extend their fuel cycle (new design of fuel al-

lowed for higher burnup) to increase the plant’s availability. In many cases the extension

of the fuel cycle coincides with the plant modernisation/modification (which required

extensive safety analysis and which were then used to justify the extension of the cycle),

which in many cases coincided with the uprate. When a fuel cycle is extended beyond

one year, the fact that there was an outage in a given year dominates the annual occupa-

tional dose. The three-year averaging tends to remove some obvious peaks (and valleys) ,

which are present in the data, but not completely, and the exact time of the outage varies

and the lengths of about 15 month present additional challenges. Moreover, if the cycle

duration was changed simultaneously with the uprate, then the comparison of the annual

occupational dose before and after the uprate is (almost) meaningless.

Recognising of the impact of uprates

During the process of data collection, some analysis of the data were undertaken to both

focus the data collection (and presentation) and make initial conclusions. Some interest-

ing patterns, as supported by the graphical presentation below, emerge:

Throughout the world, occupational doses at NPPs have been steadily decreasing over the

past decade, mainly through better application of ALARA principles, the use of better

shielding material, but also increased attention paid to occupational dose issues. The

ICRP suggested reduction in annual limits for radiation workers also impacted the overall

doses

Occupational exposure in the BWR plants is typically about 50% higher than in PWR

plants, due to specific of the design.

No direct relationship between the uprates and the occupational doses could be estab-

lished. The occupational doses on some plants seem to be higher after the uprate, while

on others seem to be lower. Without a detailed analysis (on a plant specific level) no gen-

eral conclusion could be raised.

16

As it can be seen when comparing units at multiple units sites (that are presumably oper-

ated in a same fashion) even with 3 years averaging does not remove the variations

caused by specific events

The three years averaging is helping in “smoothing” some of the obvious variations in the

annual occupational doses. Nevertheless (as discussed above) even the 3 years average

does not always allow for removing of external effects

Collective Radiation Exposure (CRE)

Ringhals2 PWR & Ringhals1 BWR

0

1000

2000

3000

4000

5000

6000

7000

1996 1998 2000 2002 2004 2006 years

man m

Sv

3-Yr CRE, BWR 3-Yr, CRE, PWR

Figure 3.1: PWR to BWR variation

Collective Radiation Exposure (CRE) Clinton 1,

uprate 2002 (20%)

0

500

1000

1500

2000

2500

3000

1996 1998 2000 2002 2004 2006 years

man

mS

V

3-Yr, CRE, Clinton2

Figure 3.2: Overall trend in occupational doses and impact of an uprate

17

Figure 3.3: Same site variation among units

Figure 3.4: Effects of the averaging on presentation

Effects of the uprate on fuel failures

The status of and the effects on fuel are one of the most important elements of the

uprates, in particular for extended uprates, where originally established fuel margins are

exceeded (requiring new design of fuel). This is the reason that the Uprate Database con-

tains information on the fuel used in each NPP evaluated. Moreover, fuel failures have a

direct negative impact on occupational doses in normal operation and accident releases.

Therefore, it is of interest to review the effect of the uprates on fuel by evaluating fuel

failures in relation to the time of uprate.

The total number of reported fuel failures since January 2000 has decreased in the US

(the trend is the same for scrams and general operational events). However, the number of

units experiencing fuel failures increased in the same period (about 80% of all units re-

ported fuel failures). The performance of fuel in PWRs and BWRs went in opposite direc-

tions, improving for PWR and deteriorating for BWR, although the failure rate (failed

Collective Radiation Exposure (CRE)

Loviisa1 (1977) & Loviisa2 (1981)

0

200

400

600

800

1000

1200

1400

1996 1998 2000 2002 2004 2006

years

man

mS

v

3-Yr CRE, Loviisa 1

3-Yr, CRE, Loviisa 2

Collective Radiation Exposure (CRE) Brunswick 2,

uprate 1996 (5%), 2002 (15%)

0

1000

2000

3000

4000

5000

6000

1991 1996 2001 2006

years

man m

SV 1-Yr CRE, Brunsw ick2

3-Yr CRE, Brunsw ick

2

18

assemblies per 1,000 installed) for PWRs is still higher than for BWRs. It appears that

whilst all fuel vendors have experienced fuel failures these failures are clustered on spe-

cific fuel models.

In PWRs, the dominant fuel failure modes are grid-to-rod fretting followed by debris-

related failures (this being greatly reduced by fuel filter and better control during outages

with open reactor). On BWRs, the dominant failures are debris fretting (five times higher

than for PWRs) and pellet-clad interaction/stress corrosion cracking (PCI-SCC). BWR

fuel designs are moving toward more closely packed fuel arrays (10 x 10), increasing the

potential for debris-induced failures. With smaller channel dimensions, the possibility of

debris-induced failures is greater. On PWRs, most failures are occurring on 11 units

(16%). Among BWRs, there are 8 units (25% of total) where most failures occur.

The debris-related failures are hard to relate to an uprate, as it depends mainly on the

operation and maintenance processes itself. The rod fretting and PCI are dependant on the

fuel loading, but higher fuel loading might be a consequence of an uprate, but also of an

extended fuel cycle. However, uprates increased thermal duty in both PWRs and BWRs.

Therefore, from a mechanistic point of view, power uprates would be likely to result in

reduced margins for fuel.

Contrary to expectation, the evaluation on the distribution of fuel failures and correlating

it with the uprates offer a highly inconclusive picture. Only 17 US units did not perform

any power uprate at all. Of those, 14 experienced at least one fuel failure. On the other

hand, 14 units performed large power uprates (12 BWR and 2 PWRs). Of those only 9

experienced a fuel failure. This suggests that there is no obvious correlation of the power

uprate with fuel failures.

Review of recent operational events related to uprates

During the process of data collection the project team initiated some limited-collection

and review of information on operational events that occurred as a consequence or are

otherwise related to uprates. The aim of this activity was to help identify any specific

aspect that could be of interest to consider either during the data collection within or on in

depth analysis on selected plants. While no specific issues were identified, some insights

of interest were noted, as below.

A review identified more than 40 events that have occurred over the past five years as a

result of inadequate design or implementation of uprates. The events involved equipment

issues, unanticipated responses to conditions, or challenges for operating staff. The num-

ber and types of events indicate that more significant consequences could occur if uprates

are not conducted in a controlled manner.

None of the events below had direct consequences on doses to the personnel or releases.

However, all of them might have contributed and/or raised the probability of inci-

dents/accidents that could have increased occupational dose or releases.

Significant aspects of events include:

− Loose parts as a result of a flow-induced, high-cycle fatigue failure on a steam

dryer cover plate

19

− Operational transients and equipment damage due to lack of training of plant staff

on changes to PCS operating characteristics

− Unanticipated challenges and degraded performance from reductions in margins

− Operation beyond licensed power levels for extended periods due to errors in

thermal power calculations following uprates

Steam Dryer Damage at a BWR

After an extended power uprate (18 %), increased steam flow rates led to a high-cycle

fatigue failure of a steam dryer cover. The plate broke into several pieces, resulting in a

10-day forced outage to retrieve the loose parts. This condition was not anticipated be-

cause the effects of the increased steam flow conditions in combination with existing

steam dome forces on the steam dryer were not well understood.

Extended Operation in an Overpower Condition of a BWR

A BWR with stretch uprate was operated at power level greater than 100 % because

changes to the process-computer calibration constants for feedwater flow were not identi-

fied when the feedwater transmitters were replaced.

Unexpected Feedwater Heater Problems at a BWR

Existing feedwater heater material condition was recognized in the preparation for a

stretch uprate, but not implemented due to budget limitations. The problem was identified

BEFORE an event occurred. 50 % of the nozzles on the feedwater heaters required repair

to mitigate the condition.

Turbine Control System Changes Result in Unanticipated Operational Challenges

at a PWR

After a stretch uprate on a PWR, operators experienced difficulty controlling turbine

speed and generator load. The need for new operating strategies was not recognized be-

fore implementation of the uprate.

Power Reduction at a PWR

Stretch uprate resulted in a reduced-stator cooling water differential-temperature operat-

ing margin. A power reduction was required to cope with the situation.

Reactor Instability in a Core after Subsequent Trip of Both Recirculation Pumps in

BWR

In parallel with the extended uprate, new fuel elements of GE11 type (9x9 fuel with part

length rods) were introduced in a small BWR4 core, thus having a mixed core of GE11

and GE8 (8x8 fuel). During the performance of stability measurements, as part of an

uprate, power oscillation was observed. Before this event the plant had not experienced

any core power oscillations.

Flow-Induced Vibration Issues (FIV issues) and steam dryer cracking

The commercial nuclear industry has experienced several incidents of steam dryer crack-

ing and FIV issues at nuclear power plants operating at extended power uprate conditions.

20

After installation of new steam dryers in two BWR units in the middle of the year, which

had an improved design to increase their structural capability, the licensee discovered

significant degradation of the electrometric relief valves (ERVs) at the end of the year.

The licensee shut down the units to repair the ERVs and restarted the units with operation

up to pre-uprate power levels.

BWR plants had operated for several years at the extended power uprate level with the

modified steam dryers without significant damage. Cracking was found later in two units.

The licensee repaired the cracks and installed additional modifications to the steam dry-

ers. The licensee plans to replace the dryers.

During outage inspection activities cracking was identified on a lower guide rod follower

bracket at the base of the steam dryer in the BWR plant, but only after several years of

operation at 5 percent power uprate conditions.

Abnormalities in Ultrasonic Flow Meter Instrumentation

Use of ultrasonic flow meter of the type used for MUR power uprates has led to unex-

pected but small differences in power level indications at some plants.

No single event listed above has any casual relation with radiological impact at affected

plants, but it does not mean that the above events could not be precursors to these events

having radiological impact. Moreover, it could be argued that some of the events (e.g.

steam dryer) contributed to occupational doses due to need for repair (in the area with

increased radiation level).

It should also be noted that most of the events are occurring at units with a power increase

of 5% or more, possibly indicating that the system interactions and PCS issues are not

always well understood or addressed during the planning or implementation of an uprate.

21

4. Analysis of the selected plants

Based on the data collected in the task #1, two PWR and two BWR plants were selected

for additional analyses.

4.1 Olkiluoto 1 and 2

4.1.1 Introduction

One of the BWR plants selected for detailed analyses was Olkiluoto nuclear power station

consisting of two twin BWR units, Olkiluoto 1 and 2 (OL1 and OL2, both reactors were

included in the review). Main reasons for that selection were:

− A considerable power uprate of 25% compared to the initial thermal power level.

− Reactor design and operation conditions very close to most of the Swedish

BWRs.

− A good availability of reactor data.

The following sections present the result of the indepth review performed for the OL1/2

plants. Data for the review had been obtained from the TVO utility owning and operating

the plants, and great acknowledgment is given to them for supplying the large amount of

information.

4.1.2 OL1/2 power uprate

4.1.2.1 Power uprate characteristics

On the west coast of Finland, in Eurajoki, Teollisuuden Voima Oy (TVO) operates two

840 MWe boiling water reactors, Olkiluoto 1 and 2 (OL1/2). Construction work began at

Olkiluoto early in 1974. The first reactor unit OL1 was supplied on a turnkey basis by the

Swedish company ASEA-ATOM AB (now Westinghouse Electric Sweden AB). In Sep-

tember 1975 construction work began on the second identical plant unit. The OL2 unit

was supplied on the same principle with the exception that TVO was responsible for the

civil construction work. The major subcontractors for the units were STAL-LAVAL Tur-

bin AB (turbine plant), ASEA AB (electrical equipment, generator), Uddcomb Sweden

AB (reactor pressure vessel), Finnatom (reactor internal parts, mechanical components),

Oy Strömberg Ab (electrical equipment) and Atomirakennus (OL1 civil construction).

The OL2 civil construction was carried out by a Finnish-Swedish consortium, Jukola. The

OL1 unit was first connected to the national grid in September 1978 and the OL2 unit in

February 1980.

The units have been uprated twice since the commissioning. The thermal power of each

reactor was increased from 2000 MW to 2160 MW in 1984 and to 2500 MW in 1998.

The corresponding nominal values of the net electrical output were 660 MWe, 710 MWe

and 840 MWe, respectively. The present study focuses on the second uprate, resulting in

22

a thermal power level of 125% compared to the initial power level. The net electrical

output from the plants during the period 1990-2006 is shown in Figure 4.1.1.

The latter uprate was a part of an extensive modernization program implemented in

1994–2006. After the modernization, the plant units fulfilled most of the safety and tech-

nical requirements for new nuclear power plants. The modernization program was in line

with TVO´s policy to keep the plant units continually up-to-date technically.

80

90

100

110

120

130

140

1990 1991 1992 1993 1994 1995 1996 1997 1998 1999 2000 2001 2002 2003 2004 2005 2006

Re

ac

tor

po

wer

[%]

OL1-Pow

OL2-Pow

Figure 4.1.1: OL1/2 – Relative reactor power based on net el. output(100% = 660 Mwe)

The reactor building (Figure 4.1.2) is the dominant and highest building of the plant. It

encloses the primary containment of the reactor and serves as a secondary containment.

The reactor service room, at the top of the building, contains the reactor and fuel pools

with storage racks for fuel and internals, the reactor service bridge for refuelling opera-

tions, and the overhead crane for handling the containment dome, the reactor vessel lid

and other heavy components. The bottom part of the reactor building contains separate

compartments for important safety-related systems, such as the emergency core cooling

systems. The reactor containment is a pre-stressed concrete vessel.

The containment is based on the principle of pressure-suppression. This allows for a

compact containment design, with the minimum of equipment installed inside the con-

tainment. The use of internal main circulation pumps has allowed further reduction of the

containment volume. All components requiring regular service during normal operation

of the reactor are located outside the containment. The tightness of the containment is

ensured by a steel liner embedded in concrete. The steel liner is protected by the concrete

against corrosion, thermal transients and hot water and steam jets or missiles that may

occur in the event of a pipe rupture. The containment is inerted, i.e. filled with nitrogen

gas during operation. The containment is divided into compartments by internal struc-

tures, the upper and lower drywell, and the wetwell. Access to the containment is gained

through air locks at the bottom of the lower drywell, and at the floor of the upper drywell.

The cylindrical part of the containment vessel extends to the top of the reactor vessel. The

condensation pool is enclosed in the annular space between the containment vessel wall

and an inner cylindrical wall, which also carries the biological shield.

23

Figure 4.1.2: OL1/2 - Section through the reactor building and the reactor containment

1 – Reactor, 2 - Main steam lines, 3 - Fuel storage pool, 4 - Reactor service bridge,

5 - Reactor service room crane, 6 - Main circulation pumps, 7 - Control rod drives,

8 - Reactor containment vessel, 9 - Control rod service platform, 10 - Blow-down pipes,

11 - Embedded steel liner. 12 - Condensation pool, 13 - Scram system tanks, 14 – Venturi scrubber

The reactor pressure vessel (Figure 4.1.3) is made of low-alloy steel, with a lining of

stainless steel. All major pipe nozzles are located above the top of the core, to ensure that

the core is kept flooded in the event of a pipe rupture in the primary systems. The reactor

vessel hangs on top of the biological shield by means of a welded-on support skirt. The

vessel support skirt is located near the primary system pipe connections, an arrangement

which minimizes the pipe stresses resulting from the thermal expansion of the vessel.

This location also allows for more maintenance space around the recirculation pumps.

The reactor internals are designed to allow for fast and safe handling during refuelling

operations. Apart from the moderator tank support skirt and the pump deck, which are

welded to the reactor vessel, all internals are removable. The internals are held in position

in the reactor vessel by means of resilient beams in the reactor vessel cover. When the

cover has been removed, the internals can be lifted out of the reactor without breaking

any bolted joints. Another related feature is the thermal insulation of the reactor cover,

which is fastened to the inside of the containment dome, they are removed together when

the reactor is to be opened. The procedures for removing the reactor vessel cover have

been considerably simplified by eliminating all the external pipe connections to the reac-

tor cover and making the connection inside the reactor.

24

Figure 4.1.3: OL1/2 - Sectional view of the reactor pressure vessel

The coolant flow through the core is maintained by means of six internal circulation

pumps. The internal circulation pump design is based on the use of wet motors, thus

eliminating shaft seals. The motor housing forms an integral part of the reactor vessel.

Internal circulation pumps offer a number of advantages over external pumps:

− no risk of major pipe rupture below the top of the core

− compact containment design

− low circulation pressure drop improves natural circulation and decreases auxiliary

power demand

− lowered drywell background radiation level contributes to very low occupational

exposure during pump motor maintenance and inspection

− significant reduction of primary system weld length.

A split shaft design allows for convenient assembly and disassembly. The pump shaft

extends into the hollow motor shaft and power is transmitted from the motor shaft

25

through a coupling that can be disassembled from the bottom of the motor housing. A

pump motor or impeller can thus be removed or replaced without draining the water from

the reactor vessel.

The turbine plant comprises a single turbine-generator unit. It is a 3000 rpm tandem-

compound, single-shaft machine with one high pressure (HP) cylinder and four low pres-

sure (LP) cylinders. The turbine is equipped with a single-pass condenser, mounted

across the longitudinal axis of the turbine. The condenser is sea water cooled, and

equipped with titanium tubing. The heating of the condensate and feedwater up to a tem-

perature of 185°C is carried out in five stages. Both the LP heaters and the HP feed heat-

ers are split up into two half-capacity, parallel circuits, each equipped with a bypass sys-

tem.

The purpose of the offgas system (Figure 4.1.4) is to limit the emission of radioactive

gases from the plant. The system employs charcoal absorption, and consists basically of

two decay vessels, two dryers, two fans and three charcoal columns. The gas from the

turbine ejectors flows through the recombiner system, the first decay vessel, one of the

dryers, one of the charcoal columns, one of the fans and finally the second decay vessel.

The xenon content in the offgas flow will be absorbed in the charcoal. When the absorp-

tion capacity of the column has been used, the flow is automatically routed through an-

other column. The “used” column is then connected to the turbine condenser, and the

xenon in its charcoal is desorbed by a small flow of “cleaned” air to the condenser.

The main difference between this offgas system and other charcoal-type systems is that it

uses a relatively small quantity of charcoal and that the radioactive gases decay in sand

beds instead of charcoal columns. Thus, only a small fraction of the radioactive gases,

essentially low energy radiation emitters, will reach the charcoal columns.

26

Figure 4.1.4: OL1/2 – Offgas system

The primary circuit of the reactor operates without chemical additives to the coolant. That

means that neither hydrogen gas1 nor zinc

2 is injected as in many other BWRs. Feedwater

chemistry corresponds to that of “neutral water”, i.e. water with very low electrical con-

ductivity. High reactor water purity contributes substantially to reliable operation of the

reactor, prevents crud deposits on the fuel rods and reduces the radioactive contamination

of the primary systems, thus ensuring better accessibility and lower occupational radia-

tion exposure.

The primary circuit water is treated by two independent, coordinated cleanup systems, the

reactor water cleanup and the condensate cleanup systems (Figure 4.1.5). The reactor

water cleanup system comprises two ion exchanger units of radial flow, bed filter type.

Each unit is capable of producing the normal cleanup flow, corresponding to 2% of the

maximum feedwater flow, so that the flow may be doubled when necessary. The flow

1 Hydrogen Water Chemistry (HWC) in order to reduce Stress Corrosion Cracking. The operation utilized in

OL1/2 without hydrogen injection is normally classified as Normal Water Chemistry (NWC).

2 Injection of Depleted Zinc Oxide (DZO) to the reactor water is applied in many BWRs in order to reduce

radiation fields.

27

rate in the reactor water cleanup system was increased after the power uprate to maintain

the 2% capacity per filter unit. The cleanup flow is generated by one pump in the shut-

down cooling system, through two regenerative heat exchangers and one cooler, to one of

the ion exchangers, and returns to the reactor via the regenerative heat exchangers. A

portion of it flows through the scram system, purging and cooling the control rod drives.

The condensate cleanup (CCU) system, connected between the first and second conden-

sate preheaters, comprises seven parallel coupled trains with deep-action, rod-type pre-

coat filters. Note that all turbine drainage is cascaded back and cleaned in the CCU sys-

tem (Figure 4.1.5).

Figure 4.1.5: OL1/2 – Reactor water and condensate cleanup system

(Reactor water: deep bed filters, condensate: pre-coat filter demineralizers)

The reactor cores of OL1 and OL2 contain 500 fuel assemblies each, which means 5

MWth per fuel assembly at present maximum power level, compared to 4 MWth per as-

sembly at initial design power. The fuel assemblies for the first cores and subsequent

reloads of OL1 and OL2 were of 8x8 design. Since then, new fuel designs have gradually

been adopted. At present 10x10 fuel is used in both reactors. This type of fuel has new

features that have made reactor power uprating and more efficient fuel utilizations possi-

ble. The linear heat load has in principle been maintained compared to the initial core.

Main data for the OL1 and OL2 reactors are provided in the Table 4.1.1.

28

Table 4.1.1: OL1/2 – Main data

A summary of all modifications that have been introduced in the OL1/2 plants during the

period 1994 – 2006 is presented in Table 4.1.2 and Table 4.1.3. Note that all modifica-

tions are not mainly due to the power uprates, but in many cases part of the ongoing plant

modernization programs carried out. Those two aspects are not always easy to separate,

introduced modifications are in many cases addressing both aspects. The major bulk of

work in association to the power uprates was performed during the 1996, 1997 and 1998

outages in OL1, and during the 1997 and 1998 outages in OL2. Note also that major mod-

ernization work was carried out during 2005 in OL2, and during 2006 in OL1. The varia-

tion in outage lengths 1990 – 2006 is presented in Figure 4.1.6. The normal strategy is

that a larger outage is performed at one of the plants during one year, but only a shorter

outage for refuelling at the other plant, and vice versa the next year. The years 1997, 1998

and 2000, however, display somewhat prolonged outages in both plants. Generally speak-

ing, both plants are characterized by short outages and high plant availability, and the

modernization and power uprate programs have been carried out with a rather small im-

pact on average outage length (impact on occupational exposures will be addressed later

in the report).

29

A summary of all modifications that have been introduced in the OL1/2 plants during the

period 1994 – 2006 is presented in Table 4.1.2 and Table 4.1.3. Note that all modifica-

tions are not necessarily due to the power uprates, but in many cases were part of the

ongoing plant modernization programs carried out. Those two aspects are not always easy

to separate; introduced modifications are in many cases addressing both aspects. The

major bulk of work in association to the power uprates was performed during the 1996,

1997 and 1998 outages in OL1, and during the 1997 and 1998 outages in OL2. Note also

that major modernization work was carried out during 2005 in OL2, and during 2006 in

OL1. The variation in outage lengths 1990 – 2006 is presented in Figure 4.1.6. The nor-

mal strategy is that a larger outage is performed at one of the plants during one year, but

only a shorter outage for refuelling at the other plant, and vice versa the following year.

The years 1997, 1998 and 2000, however, display somewhat prolonged outages in both

plants. Generally speaking, both plants are characterized by short outages and high plant

availability, and the modernization and power uprate programs have been carried out with

a rather small impact on average outage length (impact on occupational exposures will be

addressed later in the report).

Table 4.1.2: OL1/2 – Summary of modernization work performed 1994 - 1999

YEAR OL1 OL2

1994 • Replacement of two vessels in Condensate

system to SS-steel.

• Replacement of two heat exchangers in Con-

densate system.

• Replacement of core grid.

• Replacement of RHR syst. piping app. 185 m

outside of containment.

• Replacement of Main generator.

• Replacement of two vessels in Condensate

system to SS-steel.

1995 • Modifications of operating equipment of per-

sonnel air lock doors of PS-containment

• Modifications of operating equipment of

personnel air lock doors of PS-containment.

• Replacement of four heat exchangers in Con-

densate system

1996 • Installation of two new safety relief valves with

blow down piping.

• Replacement of LP3- and LP4-turbines.

• Replacement of Main generator.

• Modifications of HP-turbine.

• Modernization of turbine automation.

• Increase of capacity of Secondary cooling

systems.

• Replacement of condensate lines to SS-steel

(app. 50 m)

• Modernization of Reactor service bridge.

• Increase of power to 113% (2268 MW).

• Preparation works for the 1997 moderniza-

tion.

1997 • Installation of new Moderator tank head /

Steam separator unit.

• Replacement of 2 valves and some piping in

RHR syst.

• Replacement of some piping in the Core spray

syst.


• Modernization of Condensate pumps P1-P6 and

modifications of Condensate pipelines.

• Change of Generator switch.

• Installation of new Moderator tank head /

Steam separator unit.

• Installation of two new safety relief valves

with blow down piping.


• Modifications of HP-turbine.

• Replacement of some piping in the RWCU

syst.

• Increase of capacity of Secondary cooling

systems.

30

YEAR OL1 OL2

1998 • Modifications of HP-turbine and replacement

of shaft sealings.

• Replacement of six vessels in the Reheater and

moisture separator syst. to larger.

• Installation of new moisture separators in steam

lines to the Reheater and moisture separator

syst.

• Related to the modifications of turbine plant,

replacement of plenty of pipelines in Reheater

and moisture separator and Steam extraction

syst.

• Modernization of the reheaters in the Reheater

and moisture separator syst.

• Replacement of HP-regulator valves to larger

and changes for main steam lines.

• Lifting of level of two condensate tanks (1 m)

• Modernization of Boron syst.

• MODE-project complete. Increase of power

to 2500 MWth (840 MWe net)

• Replacement of 2 valves and some piping in

RHR syst.


• Modifications of HP-turbine and replacement

of shaft sealings.

• Replacement of six vessels in the Reheater

and moisture separator syst. to larger.

• Installation of new moisture separators in

steam lines to the Reheater and moisture sepa-

rator syst.

• Related to the modifications of turbine plant,

replacement of plenty of pipelines in Reheater

and moisture separator and Steam extraction

syst.

• Modernization of the reheaters in the Reheater

and moisture separator syst.

• Replacement of HP-regulator valves to larger

and changes for main steam lines.

• Lifting of level of two condensate tanks (1 m)

• Modernization of condensate pumps P1-P6

and modifications of condensate pipelines.

• Change of Generator switch.

• Modernization of Boron syst.

• MODE-project complete. Increase of

power to 2500 MWth (840 MWe net)

1999 • Modifications of Core spray pipelines inside of

RPV.

• Assuring the opening of two safety relief valves

during severe reactor accident conditions.

• Commutation of halon to halotron in Fire ex-

tinguisher system.

• Modifications of Core spray pipelines inside

of RPV.

• Assuring the opening of two safety relief

valves during severe reactor accident condi-

tions.

• Commutation of halon to halotron in Fire

extinguisher system.

Table 4.1.3: OL1/2 – Summary of modernization work performed 2000-2006

YEAR OL1 OL2

2000 • Replacement of two valves in the RHR syst.

and parts of the piping after these valves.

• Replacement of steam extraction pipelines from

LP1- and LP2-turbines to one LP feedwater

heater.

• Replacement of the main part of the piping of

the Flange cooling syst. in- and outside of the

containment.

• Changes for the suction lines of pumps P1 and

P2 and modification of the tank T1 of Boron

syst.

• Replacement of the pumps P1-P3 in the Tur-

bine oil syst.

• Replacement of steam extraction pipelines

from LP1- and LP2-turbines to one LP feed-

water heater.

• Changes for the suction lines of pumps P1

and P2 and modification of the tank T1 of

Boron syst.

31

2001 • Modifications introduced for lower personnel

air lock of PS-containment by considering pos-

sible steam explosion.

• Replacement of one valve in the RHR syst.

and part of the piping after the valve.

• Replacement of the main part of the piping of

the Flange cooling syst. in- and outside of the

containment.

• Install of a new service valve in Flange cool-

ing system.

• Modifications for the fluid couplings and

changes of the rotor wheels of the feedwater

pumps P301. P302 and P304.

2002 • Replacement of the water mixing location of

Feedwater and Auxiliary feedwater system.

• Install of a new service valve in Flange cooling

system.

• Modifications for the fluid couplings of the

feedwater pumps P301 and P302.

• Modifications introduced for lower personnel

air lock of PS-containment by considering

possible steam explosion.

2003 • Modifications for the fluid couplings of the

feedwater pumps P303 and P304.

• Replacement of the feedwater spargers in the

RPV.

• Condensate cleanup project, including proc-

ess modifications, replacement of heat ex-

changer E1 and increasing the pressure of

condensate pumps P5 and P6 with 3.5 bar.

2004 • Replacement of the feedwater spargers in the

RPV.

• Main steam line supports replacements.

• Condensate cleanup project, including process

modifications, replacement of heat exchanger

E1 and increasing the pressure of condensate

pumps P5 and P6 with 3.5 bar.

• Preparative works for the replacement of Mois-

ture separator / reheaters and retrofit of HP tur-

bines at 2006.


• The "old" feedwater spargers were placed

back.

• The preparative works for the replacement of

moisture separator reheaters and retrofit of

HP turbines at 2005.

2005 • Preparative works for the replacement of Mois-

ture separator / reheaters and retrofit of HP tur-

bines at 2006.


• The replacement of Steam dryer in RPV.

• The replacement of repaired feedwater

spargers.

• The replacement of HP-turbine (rotor includ-

ing blades, inner casing including stationary

blades, steam inlet piping).

• The replacement of steam Reheater / moisture

separators and drainage piping and tanks.


2006 • The replacement of Steam dryer in RPV.

• The replacement of HP-turbine (rotor including

blades, inner casing including stationary blades,

steam inlet piping).

• The replacement of steam reheater / moisture

separators and drainage piping and tanks.


• Replacement of one inner isolation valve in

RHR syst. and one valve in the Flange cooling

syst. in the containment.

• Inspections and repair of the LP4 rotor because

of the stress corrosion cracks.


• The "old" Steam Dryer were reinstalled be-

cause of repair of the "new" Steam Dryer.

32

OL1/2 - Outage lenghts

0

5

10

15

20

25

1990 1991 1992 1993 1994 1995 1996 1997 1998 1999 2000 2001 2002 2003 2004 2005 2006

Da

ys

OL1

OL2

Figure 4.1.6: OL1/2 – Outage lengths 1900 - 2006

4.1.2.2 Water chemistry and radiochemistry

The water chemistry and radiochemistry conditions in the plants before and after the

power uprates are presented a series of diagrams, Figure 4.1.7 - Figure 4.1.15. The reac-

tor power of each plant is included in the diagrams for identification of the power uprate

history.

Figure 4.1.7 shows measured concentrations of iron in the final feedwater. The efficiency

of the condensate cleanup plant (CCU) is of a large importance for the feedwater chemis-

try, and a power uprate implies a certain strain on that system due to higher flow rates.

However, an improvement of this system had already been introduced at the beginning of

the 90:ies, and good iron removal efficiency was also maintained after the power uprate.

The increase of iron in OL2 during the end of the period is deliberate because of recom-

mendations from the fuel vendor to that plant to maintain a certain excess of iron in the

fuel crud.

0

0.5

1

1.5

2

2.5

1990 1991 1992 1993 1994 1995 1996 1997 1998 1999 2000 2001 2002 2003 2004 2005 2006

Fe

[p

pb

]

0

20

40

60

80

100

120

140

Po

we

r [%

]

OL1

OL2

OL1-Pow

OL2-Pow

Figure 4.1.7: OL1/2 – Feedwater Fe before and after power uprate

33

The CCU system is, as stated, important for maintaining good water chemistry. However

it is also the major source of sulphate, one of the most detrimental impurities in the reac-

tor water from a materials integrity point of view. (Figure 4.1.8). The source of the sul-

phate is mainly from degradation of ion exchange resin, normally from the CCU system.

There are three important factors that affect the rate of degradation of CCU resin:

1. Temperature - a power uprate normally results in an increased temperature,

which is detrimental. The increase in OL1/2 was from 60°C to 70°C.

2. The quantity of hydrogen peroxide (H2O2) in the condensate. Both carry over

with the steam and the decomposition of H2O2 which are affected by a power

uprate. Normally an increase of H2O2 is experienced.

3. The total loading of iron on the resin. The higher flow rates associated with

uprates normally imply increased loading of iron on the resin. The higher loading

of iron may demand more frequent backflushing of the CCU filters, which, on the

other hand, results in an increased production of radwaste.

Much effort has been spent in the OL1/2 plants to reduce the degradation of CCU resin.

Important modifications were introduced in 2003 in OL2 and in 2004 in OL1 (Table

4.1.3), which resulted in a decrease of operation temperature for the CCU filters from

about 70°C to about 50°C. This measure has significantly improved the situation and both

plants are today basically meeting the recommended limit (WANO limit: 2 ppb).

0

2

4

6

8

10

12

14

16

18

20

1995 1996 1997 1998 1999 2000 2001 2002 2003 2004 2005 2006

SO

4 [

pp

b]

0

20

40

60

80

100

120

140

Po

we

r [%

]

OL1

OL2

OL1-Pow

OL2-Pow

WANO limit:

2 ppb

Figure 4.1.8: OL1/2 – Reactor water sulphate before and after power uprate

Some important corrosion products, Ni, Co and Cr, in the reactor water are shown in Fig-

ure 4.1.9, Figure 4.1.10 and Figure 4.1.11, respectively. No dramatic influence of the

power uprates in 1998 is seen (the reducing trend for Ni is probably due to improvements

in sampling procedures). Note, however, an increase of Cr after the 2005 outage in OL2.

This increase is most likely due to the installation of a new moisture separator in the RPV

(Table 4.1.3) with a large unfilmed stainless steel surface. Similar increases have been

seen in other plants when installing new components with large stainless steel surfaces.

34

0

0.1

0.2

0.3

0.4

0.5

0.6

0.7

0.8

0.9

1

1990 1991 1992 1993 1994 1995 1996 1997 1998 1999 2000 2001 2002 2003 2004 2005 2006

Ni

[pp

b]

0

20

40

60

80

100

120

140

Po

we

r [%

]

OL1

OL2

OL1-Pow

OL2-Pow

Figure 4.1.9 OL1/2 – Reactor water Ni before and after power uprate

0

0.01

0.02

0.03

0.04

0.05

0.06

0.07

0.08

0.09

0.1

1990 1991 1992 1993 1994 1995 1996 1997 1998 1999 2000 2001 2002 2003 2004 2005 2006

Co

[p

pb

]

0

20

40

60

80

100

120

140

Po

we

r [%

]

OL1

OL2

OL1-Pow

OL2-Pow

Figure 4.1.10: OL1/2 – Reactor water Co before and after power uprate

35

0

1

2

3

4

5

6

7

1990 1991 1992 1993 1994 1995 1996 1997 1998 1999 2000 2001 2002 2003 2004 2005 2006

Cr

[pp

b]

0

20

40

60

80

100

120

140

Po

we

r [%

]

OL1

OL2

OL1-Pow

OL2-Pow

Figure 4.1.11: OL1/2 – Reactor water Cr before and after power uprate

Radiation levels in a BWR plant are very much determined by the activated corrosion

products, e.g. Co-60, and measured reactor water concentrations of some of these are

presented in Figure 4.1.12 - Figure 4.1.15, and are commented on below:

− Co-60: The most important activated corrosion product, being responsible for the

dominant fraction of the occupational exposures in the BWR plants. The Co-60

levels in OL1/2 show some variations, but these variations are not easily corre-

lated to the power uprates. The variation in reactor water activity is larger than

the variation in Co-60 activity on primary piping, which is described in a later

section. The Co-60 reactor water concentration is affected by other factors than

the inflow of cobalt, e.g. the corrosion product composition of the fuel crud.

0

1000

2000

3000

4000

5000

6000

7000

8000

9000

10000

1990 1991 1992 1993 1994 1995 1996 1997 1998 1999 2000 2001 2002 2003 2004 2005 2006

Co

60

[B

q/k

g]

0

20

40

60

80

100

120

140

Po

we

r [%

]

OL1

OL2

OL1-Pow

OL2-Pow

Figure 4.1.12: OL1/2 – Reactor water Co-60 before and after power uprate

36

− Co-58 is due to neutron activation of Ni, and the reactor water concentration is

very much determined by the surface area of in-core Inconel in the reactor, i.e.

Inconel in fuel spacers. This is evident by studying the two sister reactors OL1

and OL2:

- OL1: Have during the period been loaded with fuel with a small fuel spacer In-

conel area.

- OL2: -1999: about 0.5 m2 per assembly, 1999 – 2003, about 0.1 m

2 per assem-

bly, 2003-: about 0.9 m2 per assembly.

The OL1/2 data show an influence of fuel design but not of power uprate.

0

5000

10000

15000

20000

25000

30000

1990 1991 1992 1993 1994 1995 1996 1997 1998 1999 2000 2001 2002 2003 2004 2005 2006

Co

58

[B

q/k

g]

0

20

40

60

80

100

120

140

Po

we

r [%

]

OL1

OL2

OL1-Pow

OL2-Pow

Figure 4.1.13: OL1/2 – Reactor water Co-58 before and after power uprate

− Mn-54 is produced by neutron activation of iron. The effect of reduced feedwater

iron inflow due to improved CCU performance is clearly visible. Recent slight

increases are deliberate, i.e. a small increase of feedwater iron is arranged in or-

der to follow fuel vendor recommendations.

0

2000

4000

6000

8000

10000

12000

14000

16000

18000

20000

1990 1991 1992 1993 1994 1995 1996 1997 1998 1999 2000 2001 2002 2003 2004 2005 2006

Mn

54

[B

q/k

g]

0

20

40

60

80

100

120

140

Po

we

r [%

]

OL1

OL2

OL1-Pow

OL2-Pow

Figure 4.1.14: OL1/2 – Reactor water Mn-54 before and after power uprate

37

− Sb-124 is produced through neutron activation of antimony, and only small quan-

tities of antimony are needed to produce a significant amount of Sb-124. The re-

sulting amount of Sb-124 is primarily determined by the inflow of antimony, but

also by the corrosion product balance in the reactor water3. A suspected main

source of antimony is antimony-containing graphite sealings in the feedwater

pumps, a material that has been shown to be not easily be replaced by the anti-

mony-free substitutes. The Sb-124 activity does not seem to be directly associ-

ated with the power uprate.

1.E+01

1.E+02

1.E+03

1.E+04

1990 1991 1992 1993 1994 1995 1996 1997 1998 1999 2000 2001 2002 2003 2004 2005 2006

Sb

12

4 [

Bq

/kg

]

0

20

40

60

80

100

120

140

Po

we

r [%

]

OL1

OL2

OL1-Pow

OL2-Pow

Figure 4.1.15: OL1/2 – Reactor water Sb-124 before and after power uprate

The amount of moisture in the steam leaving the reactor pressure vessel is supervised by

comparing measured Na-24 in condensate and reactor water. The results of such meas-

urements in OL1/2 are presented in Figure 4.1.16. A significant variation is seen through

the years:

− 1996: Some problems with enhanced steam moisture were already experience

during the period before the power uprate in 1998. The average moisture content

was, however, controlled at a level of about 0.1% through optimized fuel loading.

− 1996-1997: Some further enhancement was seen, especially in OL1 when the

power was gradually increased.

− 1997-1998: New moderator tank lids including steam separators were installed in

both plants during the 1997 outage. Some reduction in moisture was seen during

the following cycle.

− 1998-2005: The power uprate after the 1998 outage resulted in a significant en-

hancement of moisture, with typical levels of 0.3 – 0.4%. This increase resulted

3 Excess of Fe compared to Ni (+Zn) means low solubility and low reactor water concentrations, and vice

versa.

38

in a considerable buildup of radiation fields around main steam lines and other

turbine components (discussed in more detail in a later section).

− 2005-2007: New improved steam dryers were installed in the reactor pressure

vessels, 2005 in OL2 and 2006 in OL1. The new dryers turned out to have a dra-

matic effect on the steam moisture, with a reduction down to about 0.01%. A

crack in a weld in the new dryer in OL2 was observed during the 2006 outage,

which resulted in the temporary reinstallation of the old dryer and a return to the

enhanced steam moisture level during the present cycle (The crack is currently

being repaired, and the new dryer is planned to be reinstalled during the 2007

outage).

The influence on the radiation fields of the variation in steam moisture will be discussed

in a later section.

0

0.1

0.2

0.3

0.4

0.5

0.6

0.7

1990 1991 1992 1993 1994 1995 1996 1997 1998 1999 2000 2001 2002 2003 2004 2005 2006

Ste

am

mo

istu

re [

%]

0

20

40

60

80

100

120

140

Po

we

r [%

]

OL1

OL2

OL1-Pow

OL2-Pow

Figure 4.1.16: OL1/2 – Steam moisture content before and after power uprate

New steam dryers: OL1: 2006 - , OL2: 2005-2006

39

4.1.2.3 Radiation levels

4.1.2.3.1 During operation

The radiation levels during operation is mainly determined by short-lived nuclides such

as N-16 (T½ = 7.12 s). The production occurs through neutron activation of the 16

O iso-

tope in the coolant, which means that the production rate is basically proportional to the

thermal reactor power. The N-16 radiation source term in different systems is affected by

the production rate, but also by the decay time from the reactor circuit due to the N-16

nuclide’s short half-life. Power uprates affect the flow rates, and hence the decay times.

The distribution of N-16 between the steam and the reactor water is dramatically affected

by injection of hydrogen gas, so called HWC, with an increase of steam line activity to

about a factor of five. Operation without hydrogen gas injection, so called NWC, has,

however, been maintained in both OL1 and OL2, and any introduction of HWC is not

considered.

One way to assess the overall affect of power uprates on radiation levels during operation

is to study the occupational exposures obtained during reactor operation. Data from

OL1/2 are summarized in Figure 4.1.17. It must be noted, that the annual exposure dur-

ing operation is normally only a small fraction, 10-15%, of the total occupational expo-

sure, and is not always well distributed between the different reactors. The data in Figure

4.1.17 show, therefore, some variation, but no clear correlation to the 1998 power uprate

is seen.

OL1/2 - Annual occupational exposures during operation

0

20

40

60

80

100

120

140

160

1990 1991 1992 1993 1994 1995 1996 1997 1998 1999 2000 2001 2002 2003 2004 2005 2006

Exp

os

ure

[m

ma

nS

v]

OL1

OL2

Figure 4.1.17: OL1/2 – Annual occupational exposures during reactor operation before

and after power uprate

40

4.1.2.3.2 During outages

The radiation fields during outage conditions in OL1/2, as well as in other Nordic BWRs,

have been regularly followed-up. These measurement campaigns have included both dose

rate surveys as well as gamma scanning measurements revealing the contribution from

different radionuclides to the radiation fields. The variation of average contribution from

different nuclides is shown in Figure 4.1.18 and Figure 4.1.19 F for OL1 and OL2, re-

spectively. Co-60 is by far the most important nuclide, with a contribution of 65-80%.

Co-58 is the second most important with a 5-15% contribution. Note especially the dif-

ference between OL1 and OL2, which correlates well with the difference in Co-58 reactor

water activity (Figure 4.1.13). The Mn-54 contribution has been gradually reduced in

parallel to a reduced inflow of feedwater iron (Figure 4.1.7). The Sb-124 contribution

varies, but during some years it reaches about 10%. The contribution from other nuclides

(Fe-59, Cr-51, Zr-95/Nb-95, etc.) is normally <10%.

0%

20%

40%

60%

80%

100%

19

90

19

91

19

92

19

93

19

94

19

95

19

96

19

97

19

98

19

99

20

00

20

01

20

02

20

03

20

04

20

05

20

06

Other

Sb124

Mn54

Co58

Co60

Figure 4.1.18: OL1 – Contribution from different radionuclides to outage radiation levels

Average value of 12 different measurement locations

41

0%

10%

20%

30%

40%

50%

60%

70%

80%

90%

100%

19

90

19

91

19

92

19

93

19

94

19

95

19

96

19

97

19

98

19

99

20

00

20

01

20

02

20

03

20

04

20

05

20

06

Other

Sb124

Mn54

Co58

Co60

Figure 4.1.19: OL2 – Contribution from different radionuclides to outage radiation levels

Average value of 12 different measurement locations

The OL1/2 gamma scanning data are compared to corresponding data from other Nordic

BWRs in Figure 4.1.20. The general picture is quite similar in all plants.

0%

10%

20%

30%

40%

50%

60%

70%

80%

90%

100%

B1

-9

7

B2

-0

4

F1

-0

3

F2

-0

3

F3

-0

3

O1

-9

6

O2

-9

7

O3

-9

8

R1

-0

4

OL1

-0

6

OL2

-0

5

Other

Sb124

Mn54

Co58

Co60

Figure 4.1.20: Contribution from different radionuclides to outage radiation levels

Comparison of 11 Nordic BWRs

Measured dose rates in different locations at outage conditions in OL1/2 during the period

1990 – 2006 are presented below in a series of diagrams. The OL1/2 dose rates are also

compared with the database covering all eleven Nordic BWRs (denoted “BWRs”, the

general database is not updated in the same degree as the OL1/2 data).

42

Of large importance for occupational exposures are the radiation fields around primary

piping containing hot reactor water during operation. Such a standard location in the Nor-

dic BWRs is on a vertical pipe just outside the containment in the RHR system, and

measured dose rates are shown in Figure 4.1.21. The corresponding gamma scanning

results in OL1/2 is shown in Figure 4.1.22. The radiation levels display some variation,

but OL1/2 has basically remained on the same level as before the 1998 power uprate. The

OL1/2 levels are on the lower end of the database. The gamma scanning data do not show

any significant change in contribution from different radionuclides.

0

1

2

3

4

5

6

7

1990 1992 1994 1996 1998 2000 2002 2004 2006

DR

[m

Sv

/h]

BWRs

OL1

OL2

OL1/2:

MP 68, 70

Figure 4.1.21: OL1/2 – Radiation levels on vertical RHR lines compared to database

for Nordic BWRs (power uprates in OL1/2 during 1998)

43

0

0.5

1

1.5

2

2.5

3

3.5

4

1990

1991

1992

1993

1994

1995

1996

1997

1998

1999

2000

2001

2002

2003

2004

2005

2006

DR

[m

Sv

/h]

Other

Sb124

Mn54

Co58

Co60

Dose rate

0

0.5

1

1.5

2

2.5

3

1990

1991

1992

1993

1994

1995

1996

1997

1998

1999

2000

2001

2002

2003

2004

2005

2006

DR

[m

Sv

/h]

Other

Sb124

Mn54

Co58

Co60

Dose rate

Figure 4.1.22: OL1/2 – Gamma scanning data and radiation levels RHR/RWCU pipes,

average for 3 locations (power uprates in OL1/2 during 1998)

The contact dose rates on primary piping are mainly determined by the buildup of an

active oxide layer on the inside of the pipe. However, general radiation levels and dose

rates on certain components are also affected by the contribution from sedimented crud

(“hot spots”) in locations especially prone to accumulate particles (stagnant flow, low-

points, etc.).

The general radiation levels in rooms for RHR pumps are presented in Figure 4.1.23. The

corresponding general radiation levels in rooms with piping for the RWCU system are

shown in Figure 4.1.24. Dose rates on regenerative heat exchangers in the RWCU system

are presented in Figure 4.1.25 and the dose rates on coolers in the RWCU system in Fig-

ure 4.1.26. The radiation levels on water tanks in the scram system are shown in Figure

4.1.27, and the general radiation levels in the area below the motors for the control rod

drives in Figure 4.1.28. After the 1998 power uprate quite stable radiation levels were

experienced in all these locations. Some variations between the plants can be explained

by other factors. E.g. the difference in radiation levels below the control rod drives (Fig-

ure 4.1.28) between OL1 and OL2 during the 90’s is explained by the combination of

fuel failures with fuel dissolution and some flaking fuel channels in the OL2 plant in the

beginning of the 90’s (gamma scanning measurements revealed a rather high fraction of

Zr-95/Nb-95 activity). Both factors contributed to the release of insoluble fuel crud, and

the control rod drives and core instrumentation flanges below the reactor pressure vessel

are especially prone to accumulate such activity. This difference between OL1 and OL2

is also seen in some other locations.

44

0

0.1

0.2

0.3

0.4

0.5

0.6

0.7

0.8

0.9

1

1990 1992 1994 1996 1998 2000 2002 2004 2006

DR

[m

Sv

/h]

BWRs

OL1

OL2

OL1/2:

MP 10, 15

Figure 4.1.23: OL1/2 – General radiation levels in RHR pump rooms

compared to databasefor Nordic BWRs (power uprates in OL1/2 during 1998)

0

0.2

0.4

0.6

0.8

1

1.2

1.4

1.6

1.8

2

1990 1992 1994 1996 1998 2000 2002 2004 2006

DR

[m

Sv

/h]

BWRs

OL1

OL2

OL1/2:

MP 49,52

Figure 4.1.24: OL1/2 – General radiation levels in rooms with RWCU piping

compared to database for Nordic BWRs (power uprates in OL1/2 during 1998)

45

0

1

2

3

4

5

6

7

8

9

10

1990 1992 1994 1996 1998 2000 2002 2004 2006

DR

[m

Sv

/h]

BWRs

OL1

OL2

OL1/2:

MP 29

Figure 4.1.25: OL1/2 – Radiation levels on RWCU regenerative heat exchangers


0

1

2

3

4

5

6

7

8

9

10

1990 1992 1994 1996 1998 2000 2002 2004 2006

DR

[m

Sv

/h]

BWRs

OL1

OL2

OL1/2:

MP 30

Figure 4.1.26: OL1/2 – Radiation levels on RWCU coolers


46

0

0.05

0.1

0.15

0.2

0.25

0.3

1990 1992 1994 1996 1998 2000 2002 2004 2006

DR

[m

Sv

/h]

BWRs

OL1

OL2

OL1/2:

MP 20, 23

Figure 4.1.27: OL1/2 – Radiation levels on water tanks in scram system


0

0.1

0.2

0.3

0.4

0.5

0.6

1990 1992 1994 1996 1998 2000 2002 2004 2006

DR

[m

Sv

/h]

BWRs

OL1

OL2

OL1/2:

MP 118

Figure 4.1.28: OL1/2 – General radiation levels below control rod drive motors


47

A different picture is seen for the radiation levels on main steam lines (Figure 4.1.29,

Figure 4.1.30) and other turbine system components, e.g. the moisture separator / steam

reheater (Figure 4.1.31). The radiation levels especially on the main steam lines are

greatly affected by the steam moisture content (Figure 4.1.16), with a considerable in-

crease after the 1998 power uprate when the moisture was enhanced. On the other hand,

the reduced moisture content in OL2 during the cycle 2005-06 resulted in a rather quick

decrease of radiation fields during only one cycle down to about the pre-1998 levels, and

further operation with very low moisture content is expected to result in further reduc-

tions. A high contamination level of the turbine plant has a large impact on the occupa-

tional exposures, and a reduction of the turbine plant radiation fields is consequently an

important ALARA measure.

0

0.1

0.2

0.3

0.4

0.5

0.6

0.7

0.8

1990 1992 1994 1996 1998 2000 2002 2004 2006

DR

[m

Sv

/h]

BWRs

OL1

OL2

OL1/2:

MP 88, 89,

90, 91

Figure 4.1.29: OL1/2 – Radiation levels on main steam lines outside containment


48

0

0.1

0.2

0.3

0.4

0.5

0.6

1990

1991

1992

1993

1994

1995

1996

1997

1998

1999

2000

2001

2002

2003

2004

2005

2006

DR

[m

Sv

/h]

Other

Sb124

Mn54

Co58

Co60

Dose rate

0

0.05

0.1

0.15

0.2

0.25

1990

1991

1992

1993

1994

1995

1996

1997

1998

1999

2000

2001

2002

2003

2004

2005

2006

DR

[m

Sv

/h]

Other

Sb124

Mn54

Co58

Co60

Dose rate

Figure 4.1.30: OL1/2 – Gamma scanning data and radiation level on main steam line

(power uprates in OL1/2 during 1998)

0

0.02

0.04

0.06

0.08

0.1

0.12

0.14

0.16

1990 1992 1994 1996 1998 2000 2002 2004 2006

DR

[m

Sv

/h]

BWRs

OL1

OL2

OL1/2:

MP 198

Figure 4.1.31: OL1/2 – Radiation levels on moisture separators / steam reheaters


OL1

OL2

49

At the above locations radiation levels are mainly determined by activated corrosion

products, especially Co-60, with only a small contribution from fission products. How-

ever the radiation levels around the offgas system (Figure 4.1.32) are strongly dependent

on fission products, especially noble gas daughters such as Ba-140 and La-140. The

amount of noble gas daughters are to a large extent determined by the amount of tramp

uranium on the core, This is illustrated by comparison between OL1 and OL2 in Figure

4.1.32. As discussed, the OL2 plant suffered from fuel failures in the early 90’s resulting

in significant tramp uranium and a consequent increase in radiation fields around the

offgas system. The memory effect of tramp uranium on the core is considerable, and it

takes 10 years or more until the radiation levels are restored. During the 90’s OL1 had a

rather low tramp uranium contamination, and the radiation levels around the offgas sys-

tem were consequently low. Fuel failures with some tramp uranium contamination oc-

curred around 2000, resulting in an increase of radiation levels. Learning from this, the

utilities have introduced operational strategies to restrict the amount of tramp uranium. It

is, therefore, unlikely that the OL2 conditions occurring at the beginning of the 90’s will

be repeated today.

0.0001

0.001

0.01

0.1

1

10

100

1990 1992 1994 1996 1998 2000 2002 2004 2006

DR

[m

Sv

/h]

BWRs

OL1

OL2

OL1/2:

MP 177, 185

Figure 4.1.32: OL1/2 – Radiation levels on inlet pipe to offgas system


4.1.2.4 Occupational exposures

The annual occupational exposures in OL1 and OL2 during the period 1990 – 2006 are

presented in Figure 4.1.33. The exposures are divided into contribution during outage

and operation conditions. The exposures during operation are only a small fraction of the

total exposures, and have been quite constant during the period (Figure 4.1.17). A clear

correlation with the outage lengths (Figure 4.1.6) is seen, i.e. outages involving large

efforts for the power uprate and plant modernization projects have consequently implied

increased exposure (compare with Table 4.1.2 and Table 4.1.3). Data on exposure per

outage day are compiled in Figure 4.1.34. The exposure per outage day shows some

variation, but no trend of increasing exposure after the 1998 power uprate is seen.

50

OL1 - Annual occupational exposures

0

200

400

600

800

1000

1200

1400

1600

1800

2000

1990

1991

1992

1993

1994

1995

1996

1997

1998

1999

2000

2001

2002

2003

2004

2005

2006

[mm

an

Sv

]

Operation

Outage

OL2 - Annual occupational exposures

0

200

400

600

800

1000

1200

1400

1600

1800

2000

1990

1991

1992

1993

1994

1995

1996

1997

1998

1999

2000

2001

2002

2003

2004

2005

2006

[mm

an

Sv

]

Operation

Outage

Figure 4.1.33: OL1/2 – Annual occupational exposures showing proportion due to outage

and operation conditions

51

OL1/2 - Occupational exposure per outage day

0

10

20

30

40

50

60

70

80

90

100

1990 1991 1992 1993 1994 1995 1996 1997 1998 1999 2000 2001 2002 2003 2004 2005 2006

[mm

an

Sv

]

OL1

OL2

Figure 4.1.34: OL1/2 – Occupational exposures per outage day

The OL1/2 annual exposures are compared to some international BWR data in Figure

4.1.35 The OL1/2 exposures are rather low compared to the average international BWR.

Only years with considerable efforts in modernization projects, e.g. 2005 in OL2 and

2006 in OL1, result in exposures comparable with the average US and Japanese plants.

BWRs - Annual Exposures - International comparison 1990 -

0

0.5

1

1.5

2

2.5

3

3.5

4

1990 1992 1994 1996 1998 2000 2002 2004 2006

[ma

nS

v]

SWE+FIN

USA

JPN

GER

CH + ESP

OL1

OL2

Figure 4.1.35: OL1/2 – Annual occupational exposures compared to international data

4.1.2.5 Summary and conclusions

The data and experience from the OL1/2 plants before and after the 1998 power uprate

has been been reviewed and resulted in the following conclusions:

− The plants have been uprated twice since the commissioning. The thermal power

of each reactor was increased from 2000 MW to 2160 MW in 1984 and to 2500

52

MW in 1998. The present study has focused on the second uprate, resulting in a

thermal power level of 125% compared to the initial power level.

− The 1998 uprate was part of an extensive modernization program implemented in

1994–2006. After the modernization, the plants fulfil most of the safety and tech-

nical requirements for new nuclear power plants.

− It was possible to implement the modernisation program, with a reasonable im-

pact on outage lengths, due to good planning (maximum annual outage length 22

days compared to typically 7-14 days).

− The water chemistry is characterized by “as clean as possible”. Standard normal

water chemistry (NWC) without hydrogen injection is maintained and no zinc is

injected. The condensate polishing plant is operated to meet the power uprate

conditions e.g. a reduction of operation temperature. The flow rate in the RWCU

system has been increased to continue to maintain the 2% of the maximum feed-

water flow capacity after the power uprate. The overall result is that favourable

water chemistry conditions have been maintained, or improved, after the power

uprate.

− Some water chemistry transients have, however, been related to major replace-

ment of reactor components, e.g. the steam dryers. The most dominant specie in

the transients is chromium. The transients are restricted to the first cycle after the

replacement.

− Co source reduction was considered in the modernization, e.g. replacement of

Stellite in valves had been considered in cases of valve modification. An initial

Stellite area of about 94 dm2 in reactor system valves per plant has been reduced

by about 25% in OL1 and about 35% in OL2. Although the turbine plant contains

approximately 190 dm2 Stellite per unit downstream of the condensate cleanup

system the reduction of these Stellite areas has so far been limited. A moderate

overall reduction of Co inflow to the primary circuit seems to be indicated.

− The exposures during operation have been maintained at a constant and low level

since the uprate, even if an unavoidable increase of the source term of short-lived

N-16 activity has occurred. One important factor is that the plants have been

maintained on NWC, i.e. the increase of N-16 main steam line radiation con-

nected to HWC operation has been avoided.

− Radiation levels during outage of reactor systems have been maintained at a low

and constant level since the power uprate. Conversly radiation levels around main

steam lines and other turbine components have increased considerably after the

uprate in spite of the installation of new steam separators in both plants in 1997.

The reason for the increase is a considerable rise in steam moisture content after

the uprate. The problem has been addressed by the installation of new, recently

designed steam dryers in the reactor pressure vessel. As a result, steam moisture

and radiation levels have been considerably reduced.

− The large manhour efforts expended during some outages for power uprate and

plant modernization program have resulted in increased occupational exposures.

However, the exposure per outage day has been maintained at a fairly constant

level. In spite of the large efforts for power uprate and plant modernization, the

53

average annual exposures in the OL1/2 plants have been kept on a rather low

level compared to international BWR data.

4.2 Cofrentes

4.2.1 Introduction

The second of the BWR plants selected for detailed anlyses was Cofrentes (CNC). The

main reasons for this selection were:

− A considerable power uprate of 12% compared to the initial thermal power level.

− Water chemistry conditions including hydrogen injection (HWC) and the use of

depleted zinc injection (DZO). Both HWC and DZO have great impact on the ra-

diological situation, and are applied in some of the Swedish BWRs.

− A good availability of reactor data.

The following sections present the results of the indepth review performed for the CNC

plant. Data for the review has been supplied by the Iberdrola utility owner and operator of

the plants. The utility is greatly acknowledged for supplying the large amount of informa-

tion.

4.2.2 CNC power uprate


Cofrentes nuclear station (CNC) is a single unit BWR, located in Spain, on the province

of Valencia, by the Jucar River. It is 100% owned by Iberdrola, which is the main private

utility in Spain. The plant design is General Electric BWR-6, with Mark III containment,

and belongs to the group of post-TMI plants, and had already incorporated the lessons

learned from TMI into the original design.

The CNC plant has 624 fuel assemblies in the core, had an original rated power of 2894

MWt, and is currently operating at 3237 MWt (111.85%). Commercial operation started

in 1984 with 12-month cycles at rated power. Both cycle length and thermal power have

been increased. Power has been uprated in several steps:

1. 2952 MWt (102%) in cycle 4 (Apr-88)

2. 3015 MWt (104.2%) in cycle 11 (Mar-98)

3. 3184 MWt (110%) in cycle in cycle 14 (June-02)

4. 3237 MWt (111.85%) in cycle 15 (Oct-03), by taking credit of lower feedwater

flow uncertainties.

There are some tentative plans to bring the plant up to about 120%.

54

Fuel cycle length increased to 18 month in cycle 7, cycle 15 length was 20 months, cycle

16 is planned to 22 months and cycle 17 will be a 24 months cycle. Fuel assembly design

has evolved from the original 8x8 array to 9x9 in cycle 8 and 10x10 in cycle 12, which

has advanced designs suitable for more demanding operation conditions. This evolution

has resulted in mixed cores with bundles from the same supplier up to cycle 11, and from

different suppliers starting in cycle 12. The present power level means an average of 5.19

MWt per fuel assembly.

The variation of relative thermal power level 1996 – 2006 is presented in Figure 4.2.1.

CNC - Relative thermal power 1996 - 2006 (100% = 2894 MWt)

80

85

90

95

100

105

110

115

1996 1997 1998 1999 2000 2001 2002 2003 2004 2005 2006

P_

th [

%]

Cycle 10 Cycle 11 Cycle 12 Cycle 13

Cycle 14Cycle 15 Cycle 16

Figure 4.2.1: CNC – Relative reactor power with thermal output (100% = 2894 MWt)

The design of the recirculation system is shown in Figure 4.2.2. The plant has two exter-

nal recirculation loops located at the bottom region of the reactor building, inside the

containment. Each recirculation loop consists of a centrifugal pump, a flow control valve,

two shutoff valves and jetpumps. The jetpumps are located in the downcomer inside the

reactor pressure vessel (RPV), and are used to drive the circulating core coolant flow. The

use of jet pumps in the recirculation system has the advantage that only about 1/3 of the

core flow has to be withdrawn outside the reactor vessel. The suction lines of the recircu-

lation pumps are attached to the lower part of the RPV and take water from the downward

flow in the bottom of the downcomer. The water is then pumped and distributed through

a manifold to which a number of riser pipes are connected and returned to the reactor

through the jetpumps. The recirculation loop pumps are of vertical mechanical seal type.

Cooling of the pump seal and the pump motor is provided by a closed cooling water sys-

tem.

55

Figure 4.2.2: CNC – Reactor recirculation system with two pumps (B33)

The reactor water cleanup (RWCU) system at CNC is operated during startup, shutdown,

and refuelling operation as well as during normal plant operation. A flow sheet of the

RWCU system is shown in Figure 4.2.3. The main components in the system are:

− Two parallel pumps receiving hot reactor water from the recirculation loops and

the bottom of the RPV. The pumps have mechanical seals.

− One cooling line, consisting of three regenerative heat exchangers and two cool-

ers.

− Two parallel filter demineralizers. Original design flow was 2x125 gpm (2x8.0

kg/s), later increased to 2x150 gpm (2x9.5 kg/s).

The RWCU system takes out a water flow from both of the recirculation suction lines,

and from the bottom of the RPV. The RWCU lines are made of carbon steel, as opposed

to the recirculation lines of stainless steel. The water is cooled through the regenerative

heat exchangers and coolers before passing the filter demineralizer units. The cleaned

water is returning to the reactor circuit via the feedwater lines. The maximum cleanup

56

flow corresponds to about 1% of the maximum feedwater flow, and the filters are oper-

ated at a temperature of about 120°F (50°C). The filter demineralizers are of the precoat

type, and the backwashing frequency is typically once per two weeks. Resin traps are

installed both upstream and downstream of the demineralizers.

Figure 4.2.3: CNC – Simplified RWCU scheme

The concentration of corrosion products in the final feedwater is of great importance for

the activity buildup in BWR plants. This concentration is greatly determined by the de-

sign and materials selection of the turbine plant.

A flow chart showing the main turbine design at CNC is shown in Figure 4.2.4. The de-

sign is of the type “forward pumped heater drains” (FPHD), which means that the high

pressure turbine drains are not passing the condensate cleanup system. The high pressure

drains at CNC amount to about 30% of the total feedwater flow. Two parallel SALA

High Gradient Magnetic Filters are installed in order to remove some iron from the for-

ward pumped drains.

The condensate cleanup (CCU) system in CNC consists of five precoat filter demineraliz-

ers. The operation temperature of the CCU system is in the range 50-55°C.

The condenser tubes at CNC were originally of Admirality brass, but in RFO-12 (Sep-

tember 2000) were replaced with titanium tubes. The steam extraction lines were origi-

nally of carbon steel, which accounted for a rather high erosion corrosion rate and, thus,

elevated levels of iron in the forward pumped drains (somewhat reduced through the

SALA filters). An extensive replacement of the steam extraction pipes to low-alloy steel

has been carried out, which has dramatically reduced the forward pumped heater drains

57

contribution of feedwater iron. As in most other BWRs, Stellite is present in several valve

applications both in the reactor and turbine system. The control rods in CNC were origi-

nally of the type with “pins and rollers” of Stellite. A gradual replacement of old control

rods has been performed, and the new rods are without Stellite.

Figure 4.2.4: CNC – Main balance of plant flow diagram

The main power uprate of the plant was in 2002, but the necessary plant system and op-

eration modifications have been carried out over approximately 10 years:

− July 1996: Start injection of zinc (depleted zinc oxide – DZO). The feedwater al-

ready contained earlier significant amounts of zinc due to the brass condenser

tubes (further discussed below).

− March 1997: Start operation of hydrogen injection to obtain Hydrogen Water

Chemistry (HWC) conditions. However, the initial effect of the hydrogen injec-

tion on corrosion potential (ECP) was limited, due to considerable amount of re-

actor water copper.

− September 1997 (RFO-10): Turbine moisture separator and reheater tube sheets

were replaced.

58

− September 2000 (RFO-12): Condenser revamping to titanium. The reason for the

modification was twofold; (1) reduction of the reactor water copper and (2) an in-

crease of the condenser capacity to allow the future power uprates.

− March 2002 (RFO-13): Low pressure turbine steam paths modified for power

uprate. Heater drain pumps replaced for power uprate. Feedwater pump turbines

modified for power uprate. Feedwater flow measuring system type LEFM in-

stalled. Modification in one of the condensate filter vessels to install pleated filter

elements. Partial decontamination of the RWCU system and the lower sections of

the recirculation loops (the original planned closed loop decontamination could

not be performed due to problems with the recirculation suction nozzle plugs).

− September 2003 (RFO-14): High pressure turbine steam path modifications for

power uprate. Heater drains flow control valves modified for power uprate.

− May 2005 (RFO-15): Successful decontamination of recirculation loops, RWCU

and RHR system (for dose control, not directly for power uprate). Replacement

of RWCU heat exchangers and RWCU pump internals (not directly for power

uprate). Steam dryer inspection resulted in two cracks being found in banks

cover. Weld repair had to be done in the reactor pool by divers. The defects had

not affected the moisture content, but the defect could be related to flow induced

vibrations and may, therefore, be related to the earlier power uprates.

The power uprates in CNC has included a gradual increase of reactor pressure (Figure

4.2.5) and reactor water temperature (Figure 4.2.6). The increase of reactor pressure has

implied that the steam velocity in the top of the RPV and in the main steam lines has not

been affected very much by the power uprate. This is a difference when compared to the

Scandinavian BWRs, where power uprates have been carried out with maintained reactor

pressure level, i.e. the steam velocity has been increased in the same proportion as the

power uprate.

CNC - Reactor pressure 1996 - 2004 compared to reactor power

69

70

71

72

73

74

75

1996 1997 1998 1999 2000 2001 2002 2003 2004

Pre

ss

ure

[b

ar]

80

85

90

95

100

105

110

115

P_

th [

%]

Pressure

P_th

Figure 4.2.5: CNC – Reactor pressure 1996 - 2004

59

CNC - Reactor water temperature 1998 - 2004 compared to reactor power

270

272

274

276

278

280

282

284

1998 1999 2000 2001 2002 2003 2004

Te

mp

[C

]

80

85

90

95

100

105

110

115

P_

th [

%]

Temp Loop-A

Temp Loop-B

P_th

Figure 4.2.6: CNC – Temperature in recirculation loops (A and B) 1998 - 2004

The evolution of refuelling outage lengths is shown in Figure 4.2.7. When compared to

earlier outages, the 2002 RFO-13, where most modifications for the power uprate were

introduced, does not actually differ in length. Note that the extra long outage in 2005 was

affected by an external side corrosion problem that occured in the CRD piping in the ves-

sel pedestal penetrations that caused the extension of the RFO-15 for more than 30 days

and gave some extra exposure.

CNC - Refueling outage lenghts 1996 - 2006

0

10

20

30

40

50

60

70

80

90

RFO-09 RFO-10 RFO-11 RFO-12 RFO-13 RFO-14 RFO-15

Ou

tag

e l

en

gh

t [d

]

1996 1997 1999 2000 2002 2003 2005

Figure 4.2.7: CNC – Outage lengths 1996 - 2006


The CNC feedwater chemistry is characterised by periodically high levels of iron (Figure

4.2.8), copper (Figure 4.2.9) and zinc (Figure 4.2.10). The amount of iron shows some

60

variation, but the average level has not been significantly changed during the last 10 years

when power uprates have occurred. The earlier quite high contribution from the forward

pumped heater drains has been dramatically reduced by the replacement of carbon steel

steam extraction lines with low-alloy steel pipes. However, this decrease has been com-

pensated for by an increase of iron inflow to the condensate polishing plant (10-15 ppb Fe

in the mid-90’s, presently increased to 30-35 ppb Fe), and today’s feedwater iron is

mainly determined by the leakage through the filter demineralizers. Efforts have been

made to increase the cleanup efficiency of the condensate polishing plant, but the optimi-

zation of the polishing plant is a balance between cleanup efficiency and long run-lengths

to obtain low waste production. The present level of feedwater iron is somewhat high

compared to recent water chemistry guidelines4.

CNC - Feedwater iron 1996 - 2006 compared to reactor power

0.01

0.1

1

10

1996 1997 1998 1999 2000 2001 2002 2003 2004 2005 2006

Fe

[p

pb

]

85

90

95

100

105

110

115

P_

th [

%]

Fe

P_th

Figure 4.2.8: CNC – Feedwater iron before and after power uprate

The feedwater copper historically has been rather high, but a considerable decrease is

seen after the September 2000 condenser tube replacement of brass tubes to titanium. The

reduction of feedwater zinc due to the change of condenser tubes is met by an increase of

injected zinc, i.e. the feedwater zinc level has basically been maintained or slightly has

been increased, but the natural zinc from the brass condenser tubes has been replaced

with DZO with low level of 64

Zn5. The injection of feedwater zinc is controlled so that a

target level of reactor water zinc is maintained.

4 EPRI water chemistry guidelines, 2004 revision, specifies 0.1-1.0 ppb feedwater iron for plants on HWC.

5 DZO contains about 1% of 64Zn, compared to about 50% in natural zinc. Neutron irradiation of 64Zn results

in production of the radioactive nuclide 65Zn.

61

CNC - Feedwater copper 1996 - 2006 compared to reactor power

0.001

0.01

0.1

1

1996 1997 1998 1999 2000 2001 2002 2003 2004 2005 2006

Cu

[p

pb

]

85

90

95

100

105

110

115

P_

th [

%]

Cu

P_th

Figure 4.2.9: CNC – Feedwater copper before and after power uprate

CNC - Feedwater zinc 1996 - 2006 compared to reactor power

0.01

0.1

1

10

1996 1997 1998 1999 2000 2001 2002 2003 2004 2005 2006

Zn

[p

pb

]

85

90

95

100

105

110

115

P_

th [

%]

Zn

P_th

Figure 4.2.10: CNC – Feedwater zinc before and after power uprate

The feedwater dissolved oxygen (Figure 4.2.11) is control within the range 20-200 ppb,

where the erosion-corrosion of carbon steel is maintained at a reasonable level, and the

risk of environmental cracking is low. Hydrogen injection started in March 1997, but, in

the present study, feedwater dissolved hydrogen (DH) data has only been available from

the end of 1998 (Figure 4.2.12). During most of the period with hydrogen injection, the

feedwater DH has been kept at 1 ppm, but during 2000 and in a recent fuel cycle it was

increased to about 1.67 ppm. The combination 1 ppm of feedwater DH and significant

remaining amounts of copper in the reactor water has probably not resulted in a consider-

able decrease of ECP in the recirculation loops, and especially not for the reactor inter-

nals. The gradual decrease of reactor water copper, particularly when the hydrogen injec-

tion was increased during the recent cycle meant that ECP levels were sufficiently low for

protection of the reactor materials and has probably been obtained both in the recircula-

tion lines and the bottom of the RPV.

62

CNC - Feedwater dissolved oxygen 1996 - 2006 compared to reactor power

0

20

40

60

80

100

120

140

160

180

200

1996 1997 1998 1999 2000 2001 2002 2003 2004 2005 2006

DO

[p

pb

]

85

90

95

100

105

110

115

P_

th [

%]

DO

P_th

Figure 4.2.11: CNC – Feedwater dissolved oxygen before and after power uprate

CNC - Feedwater dissolved hydrogen 1996 - 2006 compared to reactor power

0

0.2

0.4

0.6

0.8

1

1.2

1.4

1.6

1.8

2

1996 1997 1998 1999 2000 2001 2002 2003 2004 2005 2006

DH

[p

pm

]

85

90

95

100

105

110

115

P_

th [

%]

DH

P_th

Figure 4.2.12: CNC – Feedwater dissolved hydrogen (injected) before and after power

uprate (Note: HWC operation started already March 1997, see Figure 4.2.14)

63

As mentioned, the CCU system is of importance to maintain good water chemistry, but

ironically, is also the major source of one of the most detrimental impurities in the reactor

water from materials integrity point of view, sulphate. The source of the sulphate is

mainly degradation of ion exchange resin, probably from the CCU system. There are

three important factors that affect the rate of degradation of CCU resin:

1. The temperature, a power uprate normally results in an increased temperature,

which is detrimental.

2. The content of hydrogen peroxide (H2O2) in the condensate. Carry-over with the

steam and decomposition of the H2O2 are affected by a power uprate, normally an

increase of H2O2 is experienced.

3. The total loading of iron in the resin. The higher flow rates in connection with

uprates normally imply increased loading of iron on the resin. The higher loading

of iron may demand more frequent backflushing of the CCU filters, which on the

other hand results in an increased production of radwaste.

To a large extent this type of resin degradation has been avoided in the CNC plant,

mainly through the relatively low temperature in the CCU plant. Low levels of anionic

impurities in the reactor water are exhibited by low reactor water conductivity (Figure

4.2.13).

CNC - Reactor water conductivity 1996 - 2006 compared to reactor power

0

0.05

0.1

0.15

0.2

0.25

0.3

1996 1997 1998 1999 2000 2001 2002 2003 2004 2005 2006

Co

nd

[µµ µµ

S/c

m]

85

90

95

100

105

110

115

P_

th [

%]

Cond

P_th

Figure 4.2.13: CNC – Reactor water conductivity before and after power uprate

Measured reactor water dissolved oxygen (DO) is presented in Figure 4.2.14. Low ECP

levels demand DO levels ≤1 ppb, and it is only during the recent cycle that sufficiently

low oxygen levels have been achieved. However, it must, be noted that the measurement

of very low oxygen levels may be disturbed by small intrusion of air, e.g. if the measuring

equipment is connected with plastic tubing.

64

CNC - Reactor water dissolved oxygen 1996 - 2006 compared to reactor power

0.01

0.1

1

10

100

1000

1996 1997 1998 1999 2000 2001 2002 2003 2004 2005 2006

DO

[p

pb

]

80

85

90

95

100

105

110

115

120

P_

th [

%]

DO

P_th

Figure 4.2.14: CNC – Reactor water dissolved oxygen before and after power uprate

Some important corrosion products, copper, zinc and cobalt, in the reactor water are

shown in Figure 4.2.15, Figure 4.2.16 and Figure 4.2.17, respectively. The reactor water

copper is gradually reduced after the condenser retubing, but it takes some years until

sub-ppb levels are reached. A further delayed effect of the previous copper chemistry is

that system oxides contain some remaining copper. The successful decontamination cam-

paign in 2005 eliminated this effect for the recirculation loops.

CNC - Reactor water copper 1996 - 2006 compared to reactor power

0

1

2

3

4

5

6

7

8

9

10

1996 1997 1998 1999 2000 2001 2002 2003 2004 2005 2006

Cu

[p

pb

]

85

90

95

100

105

110

115

P_

th [

%]

Cu

P_th

Figure 4.2.15: CNC – Reactor water copper before and after power uprate

(cfr. Feedwater Cu, see Figure 4.2.9)

The reactor water zinc was maintained at typically ≤5 ppb before the 2003 RFO, but in-

creased to ≥5 ppb after this outage. The reason for this change was the increased radiation

level experienced during the 2003 RFO, and on recommendation issued by EPRI.

65

CNC - Reactor water zinc 1996 - 2006 compared to reactor power

0

2

4

6

8

10

12

1996 1997 1998 1999 2000 2001 2002 2003 2004 2005 2006

Zn

[p

pb

]

85

90

95

100

105

110

115

P_

th [

%]

Zn

P_th

Figure 4.2.16: CNC – Reactor water zinc before and after power uprate

(cfr. Feedwater Zn, see Figure 4.2.10)

The measured reactor water cobalt has been biased by cobalt containing Stellite in the

sampling system valves. Improved cobalt analysing procedures were introduced in 2003,

and the sampling valves were replaced with Stellite-free in 2005, and more realistic reac-

tor water cobalt concentrations, about 0.02 ppb are recorded.

CNC - Reactor water cobalt 1996 - 2006 compared to reactor power

0.001

0.01

0.1

1

1996 1997 1998 1999 2000 2001 2002 2003 2004 2005 2006

Co

[p

pb

]

85

90

95

100

105

110

115

P_

th [

%]

Co

P_th

Figure 4.2.17: CNC – Reactor water cobalt before and after power uprate

Radiation levels in a BWR plant is very much determined by the activated corrosion

products, e.g. Co-60, and measured reactor water concentrations of some of these are

presented in Figure 4.2.18- Figure 4.2.21, and are commented below:

− Co-60: The most important activated corrosion product, being responsible for the

dominant fraction of the occupational exposures in the BWR plants. The Co-60

levels in CNC show some variations, but these variations are not easily correlated

66

to the power uprates. The general trend in CNC is gradually decreasing reactor

water Co-60 concentration.

CNC - Reactor water Co-60 1996 - 2006 compared to reactor power

1.E+02

1.E+03

1.E+04

1.E+05

1996 1997 1998 1999 2000 2001 2002 2003 2004 2005 2006

Co

-60

[B

q/k

g]

85

90

95

100

105

110

115

P_

th [

%]

Co-60

P_th

Figure 4.2.18: CNC – Reactor water Co-60 before and after power uprate

− Co-58 is due to neutron activation of Ni, and the reactor water concentration is

very much determined by the surface area of in-core Inconel in the reactor, i.e.

Inconel in fuel spacers. The CNC Co-58 levels have been maintained at a fairly

constant level during the last 10 years.

CNC - Reactor water Co-58 1996 - 2006 compared to reactor power

1.E+02

1.E+03

1.E+04

1.E+05

1996 1997 1998 1999 2000 2001 2002 2003 2004 2005 2006

Co

-58

[B

q/k

g]

85

90

95

100

105

110

115P

_th

[%

]

Co-58

P_th

Figure 4.2.19: CNC – Reactor water Co-58 before and after power uprate

− Mn-54 is produced by neutron activation of iron. The CNC reactor water Mn-54

activity shows a rather large scattering, enhanced by a relatively large fraction of

insoluble activity. A clear trend is not evident; the average level has been kept at

about the same level during the last 10 years.

67

CNC - Reactor water Mn-54 1996 - 2006 compared to reactor power

1.E+02

1.E+03

1.E+04

1.E+05

1996 1997 1998 1999 2000 2001 2002 2003 2004 2005 2006

Mn

-54

[B

q/k

g]

85

90

95

100

105

110

115

P_

th [

%]

Mn-54

P_th

Figure 4.2.20: CNC – Reactor water Mn-54 before and after power uprate

− Reactor water Zn-65 is shows a decreasing trend after the retubing of the turbine

condenser, i.e. when most of the feedwater zinc has been replaced with DZO with

low content of 64

Zn.

CNC - Reactor water Zn-65 1996 - 2006 compared to reactor power

1.E+02

1.E+03

1.E+04

1.E+05

1996 1997 1998 1999 2000 2001 2002 2003 2004 2005 2006

Zn

-65

[B

q/k

g]

85

90

95

100

105

110

115

P_

th [

%]

Zn-65

P_th

Figure 4.2.21: CNC – Reactor water Zn-65 before and after power uprate

Occasional fuel failures have been experienced in CNC, e.g. during 2004 and 2006 (I-131

activity in the reactor water are presented in Figure 4.2.22). The fuel leakages have nor-

mally been caused by debris fretting, and are not judged to be related to the power

uprates.

68

CNC - Reactor water I-131 1996 - 2006 compared to reactor power

1.E+03

1.E+04

1.E+05

1.E+06

1996 1997 1998 1999 2000 2001 2002 2003 2004 2005 2006

I-1

31

[B

q/k

g]

85

90

95

100

105

110

115

P_

th [

%]

I-131

P_th

Figure 4.2.22: CNC – Reactor water I-131 before and after power uprate

The moisture content in the steam in the CNC plant, both before and after the power

uprates, has been maintained on a low level, with very low contamination levels of the

turbine plant. Recent level of steam moisture content is typically 0.02%.



The radiation levels during operation are mainly determined by short-lived nuclides such

as N-16 (T½ = 7.12 s). The production occurs through neutron activation of the 16

O iso-

tope in the coolant, which means that the production rate is basically proportional to the

thermal reactor power. The N-16 radiation source term in different systems is affected by

the production rate, and also by the decay time from the reactor circuit due to the N-16

nuclide’s short half-life. Power uprates affect the flow rates, and hence the decay times.

The distribution of N-16 between the steam and the reactor water is dramatically affected

by injection of hydrogen, so called HWC, with an increase of steam line to about a factor

of five. CNC has been operating with HWC since March 1997.

Measuring campaigns of radiation levels at a great number of locations were carried out

at four different power levels, 92%, 104.2%, 107% and 110%, in April 2002. The corre-

sponding dose rates were measured at 111.8% power level in November 2003, i.e. at the

present maximum reactor power level. All measurements were performed at HWC condi-

tions with 1 ppm of feedwater DH.

The radiation levels outside main steam lines at the April 2002 measuring campaigns are

summarized in Table 4.2.1. The radiation levels are increased with average 23.8% be-

tween 92% and 110%, i.e. for a 19.6% power increase. The increase of radiation levels

seems to be very close to the power increase.

69

Table 4.2.1: CNC – Main Steam Line Radiation (MSLR) monitoring April 2002

at four different power levels and standard HWC conditions

92%

Pow.

104,2%

Pow.

107%

Pow.

110%

Pow.

∆ 110%/

92%

mSv/h 1 ppm

H 2

1 ppm

H 2

1 ppm

H 2

1 ppm

H 2

(∆Pow.

19.6%)

Steam line A 17.60 20.00 21.40 22.00 25%

Steam line B 17.10 19.40 20.40 21.10 23%

Steam line C 18.00 20.30 21.30 22.30 24%

Steam line D 19.00 21.80 22.80 23.40 23%

Av. line A-D 17.93 20.38 21.48 22.20 +23,8%

Dose rate measurements, inside and outside the plant buildings, for five power levels

between 92% and 111.8% have been assessed and are presented in Figure 4.2.23 and

Figure 4.2.24. The different locations show some variation. The highest power level,

111.8%, was measured on a later occasion when the operating conditions differed. It

shows lower radiation levels than the four previous power levels. Because of this, these

readings were excluded and the evaluation was based on the four remaining power levels

measured close to each other. The average relative increase between 92% and 110% is

close to 50%, i.e. somewhat higher than experienced for the main steam lines. Whilst the

measured radiation levels vary by almost 6 orders of magnitude, the average relative in-

crease is similar for different radiation levels, see Figure 4.2.24.

The annual occupational exposures 1990 – 2006 obtained during reactor operation are

presented in Figure 4.2.25. The exposure level has been quite stable over the last ten

years, and seems not to have be significantly affected by either the change to HWC in

1997 or the recent power uprates. The overall conclusion must be that the N-16 radiation

source term is not the dominant contributor to the occupational exposure during opera-

tion.

70

CNC: Relative radiation levels for different power levels

Normalization: 100% at P = 104.2%, totally 115 locations

10%

100%

1000%

90% 95% 100% 105% 110% 115%

Power level [%]

Re

l. r

ad

iati

on

vs P

=1

00

%

Median

Average

Figure 4.2.23: CNC – Measured relative radiation levels at five different power levels

Normalization: 100% at P=104.2%, totally 115 measuring locations, P=92%, 104:2%,

107% and 110% measured at April 2002, P=111.8% at November 2003

CNC: Rel. radiation (110%/104%) vs. radiation level (104%)

10%

100%

1000%

0.01 0.1 1 10 100 1000 10000 100000

Radiation levelP=104% [µµµµSv/h]

Re

l. r

ad

iati

on

(11

0%

/104

%)

Least

square fit

Figure 4.2.24: CNC – Relative increase of measured radiation levels between P=104.2%

and P=110% versus radiation level at P=104.2%(Totally 115 measuring locations both

inside and outside plant buildings)

71

CNC - Annual occupational exposures during operation

0

200

400

600

800

1000

1200

1400

1600

1990 1991 1992 1993 1994 1995 1996 1997 1998 1999 2000 2001 2002 2003 2004 2005 2006

Exp

os

ure

[m

ma

nS

v]

Figure 4.2.25: CNC – Annual occupational exposure during reactor operation


About 70% of the occupational exposure in CNC is received during refuelling outages,

and about 70% of this exposure comes from the drywell area. This area contains mainly

the recirculation loops with pipes, pumps and valves. The area is accessible only during

outages.

The radiation levels on the recirculation loops are surveyed in great detail during outages

due to the loops large impact on the occupational exposures. The different measuring

locations are shown in Figure 4.2.26, and some of the locations especially addressed in

the present study are indicated in the figure. Note that the decontamination performed at

the 2002 RFO only included the lower parts of the loops due to problems with some plugs

at the RPV nozzles.

72

Figure 4.2.26: CNC: BRAC measuring locations on recirculation loop A and B Average

of points 2/8 and 15/21 for loop A and B, respectively, used as reference in Figure 4.2.27

Point 4 (decontaminated) compared to point 15 (not decontaminated) in Figure 4.2.28

The vertical sections on the recirculation loops are used as reference locations for the

radiation surveys, and the historical dose rate data on these locations are shown in Figure

4.2.27. These locations were not included in the 2002 RFO decontamination campaign.

Dose rates on RWCU piping are also included in the figure. Note that the RWCU piping

is of carbon steel, while the recirculation loops are of stainless steel. The radiation levels

were quite stable up to the 2002 RFO, but a considerable increase was experienced espe-

cially on the recirculation lines at the following outages. Notably was that this consider-

able increase was not associated with any dramatic change in measured water chemistry

conditions, e.g. reactor water Co-60 (see Figure 4.2.18). Also notable was that the in-

crease was most pronounced in locations on the recirculation lines that were not affected

by the 2002 RFO decontamination campaign, and that decontaminated pipes in both the

recirculation loops (Figure 4.2.28) and the RWCU system showed a more reasonable

recontamination rate.

Large efforts have been spent to understand the post-RFO-13 behaviour in CNC. The

analysis has identified, that a key factor has been the evolution of HWC conditions due to

the gradual decrease of reactor water copper (Figure 4.2.15). The sub-ppb levels of reac-

tor water copper reached during cycle 14 meant that at least decontaminated pipes were

reaching low ECP levels, while probably non-decontaminated pipes with remaining cop-

per in the oxide were staying on a somewhat higher ECP level promoting precipitation

and incorporation of soluble corrosion products in the oxide layer. This radiation problem

has been solved in the following way:

− A successful decontamination was performed in the 2005 RFO, including all

parts of the recirculation loops.

− The hydrogen injection has been increased to assure more stable reducing condi-

tions in the recirculation loops and in the bottom of the RPV (Figure 4.2.12).

73

− The zinc injection has been increased after the 2003 RFO to obtain somewhat

higher reactor water zinc levels (Figure 4.2.16).

The implemented actions seem to have been quite successful, and the recontamination of

the recirculation lines is quite slow (Figure 4.2.27), measurements were performed at an

extra outage during 2006). The RWCU pipes of carbon steel, on the other hand, show a

more rapid recontamination. The activity buildup in the CNC plant seems to be primarily

affected by the water chemistry conditions, and in a very small degree by the power

uprating.

CNC: Dose rates on PLR and RWCU lines at different outages(RFO - Refueling outage, XO - Extra outage, decon - after decontamination)

0

2

4

6

8

10

12

14

16

1991

, RFO

-6

1992

, XO

1993

, RFO

-7

1993

, XO

1994

, RFO

-8

1996

, RFO

-9

1997

, RFO

-10

1999

, RFO

-11

1999

, XO

2000

, RFO

-12

2001

, XO

2002

, RFO

-13

2002

, RFO

-13

(dec

on)

2002

, XO

2003

, RFO

-14

2004

, XO

2005

, RFO

-15

2005

, RFO

-15

(dec

on)

2006

, XO

DR

[m

Sv/h

]

PLR-A

PLR-B

RWCU

Figure 4.2.27: CNC: Measured dose rates on recirculation lines (PLR) and RWCU lines

at different outages 1991 – 2006 (PLR-A and -B: average 2 locations (Figure 4.2.26),

RWCU: average several locations)

74

CNC: Dose rate on recirculation piping

0

2

4

6

8

10

12

14

16

18

20

RFO 13 - before decontamination RFO 13 - after decontamination RFO 14

DR

[m

Sv/h

]BRAC_4

BRAC_15

Figure 4.2.28: CNC: Dose rate on two different locations on recirculation lines

during RFO-13 (2002, before and after decontamination) and RFO-14 (2003)

BRAC_4: Affected by decontamination, BRAC_15: Not affected (Figure 4.2.26)


The annual occupational exposures in CNC during the period 1990 – 2006, have been

divided into contributions during reactor operation and outage conditions, are presented

in Figure 4.2.29. Some years are without RFO in CNC, which means a low annual occu-

pational exposure. Therefore, the rolling 3-year average has been included in the diagram

to obtain information about the long-term exposure trend. The exposure levels have been

quite stable during the 10 last years, but with a slight increasing trend. The last three

RFOs have been associated with somewhat higher exposure levels due to the combined

effect of large modification and maintenance efforts (2002, 2006), and the above dis-

cussed increased radiation levels after the 2002 RFO. The introduced measures to reduce

radiation fields are likely to imply somewhat lower future exposure levels, at least if

maintenance efforts are kept on a reasonable level.

75

CNC - Annual occupational exposures

0

1000

2000

3000

4000

5000

6000

1990 1991 1992 1993 1994 1995 1996 1997 1998 1999 2000 2001 2002 2003 2004 2005 2006

[mm

an

Sv

]Operation

Outage

3-year average

Figure 4.2.29: CNC – Annual occupational exposures split-up on outage and operation

conditions plus rolling 3 year-average based on total annual exposure

The CNC annual occupational exposures, presented as rolling 3-years average, are com-

pared with some international exposure data in Figure 4.2.30. The CNC data are quite

similar to the US BWR experience, which is understandable due to the similar reactor

designs. However, as discussed above, the slightly increasing trend seen in CNC during

last 10 years is contrary to the US experience, where a slightly decreasing trend is ob-

served. A similar increasing trend to CNC is also seen in Japanese plants. The main rea-

son for the Scandinavian and German plants being about half of the CNC exposure level

is probably due to the fact, that most of these plants are of a different design with internal

recirculation pumps, i.e. without the recirculation loops responsible for a large fraction of

the CNC exposures.

BWRs - Annual Exposures - International comparison 1990 -

0

0.5

1

1.5

2

2.5

3

3.5

4

1990 1992 1994 1996 1998 2000 2002 2004 2006

[ma

nS

v]

SWE+FIN

USA

JPN

GER

CH+ESP

CNC

3-year

average

Figure 4.2.30: CNC – Annual occupational exposures compared to international data

Note: the CNC data are rolling 3-year average

76


A review of data and experience from the CNC plant before and after the power uprates

1998-2003 has been performed. The review has resulted in the following conclusions:

− The present power level corresponds to 111.8% of the initial thermal power level,

which means on average 5.19 MWt per fuel assembly. The main power increase

was introduced in 2002, when the power level was increased from 104.2% to

110%.

− The fuel cycle lengths have been increased in parallel to the power uprates, ini-

tially from 12 months then up to 22 months for the present fuel cycle 16. Fuel as-

sembly design has evolved from the original 8x8 array to 9x9 in cycle 8 and

10x10 in cycle 12, which has necessitated more advanced designs for the more

demanding recent operation conditions.

− The modifications introduced have mainly affected the turbine plant:

- Turbine moisture separator and reheater tube sheets

- Condenser revamping, also including a change of tube material from brass to

titanium

- Low pressure turbine steam paths

- Turbine heater drain pumps

- Feedwater pump turbines

- Feedwater flow measuring system

- High pressure turbine steam path

- Heater drains flow control valves.

Due to low contamination level of the turbine plant it was possible to carry out

these modifications with a low exposure.

− The main change on the reactor side is that the reactor pressure and reactor tem-

perature have been slightly increased. This increase means that the steam velocity

in the reactor and in the main steam lines has been only marginally affected by

the power uprates.

− The reactor water chemistry is strongly determined by the design and materials

selection of the turbine plant. The turbine design with forward pumped heater

drains and initially carbon steel in most of the piping, resulted in a rather high in-

flow of iron. Most of the steam extraction pipes have currently been replaced

with low-alloy steel pipes which has considerably reduced the forward pumped

heater drain contribution to the feedwater iron, however, a significant contribu-

tion passing the condensate polishing plant still remains. The earlier brass mate-

rial in the condenser tubes resulted in a rather high inflow of copper and zinc, but

this inflow has been eliminated by changing to titanium tubes. Though a high in-

flow of zinc has been maintained by the introduction of zinc injection in order to

control radiation fields.

− Hydrogen water chemistry (HWC) was introduced in 1997. However, the reduc-

tion of corrosion potential was small due to high levels of reactor water copper.

The gradual reduction of reactor water copper, and an increase in the amount of

injected hydrogen have most likely resulted in low corrosion potentials meeting

the HWC requirements during recent cycles.

77

− The radiation levels during operation in different locations both inside and out-

side of plant buildings are affected by the production and distribution of the

short-lived nuclide N-16. Surveys at power levels in the range of 92% - 111.8%

indicate an average effect of the power increase in the order of +15% - +30%.

Such increase is reasonable considering the combined effects of higher N-16 pro-

duction rate and shorter transport times due to the increased flow rates. Of larger

importance for the general radiation levels during operation was probably the in-

troduction of HWC in 1997, that resulted in an increase of approximately a five

times in carryover of N-16 activity with the steam to the turbine plant. However,

the annual operational exposures have been maintained at a relatively stable level

during the last 10 years and do not seem to be significantly affected by either the

change to HWC in 1997 or the recent power uprates. The overall conclusion must

be that the N-16 radiation source term is not the dominant contributor to occupa-

tional exposure during operation.

− Radiation levels during outage conditions are of importance for occupational ex-

posures and are very much determined by the recirculation loops and the RWCU

piping. A considerable increase, particularly in the radiation fields around recir-

culation loops, was experienced at the 2003 refuelling outage. The increase does

not seem to be due to the power uprate, but rather to the combined water chemis-

try effect of gradually decreasing reactor water copper and HWC operation, re-

sulting in the restructuring of the oxide layers inside the recirculation loops. Sev-

eral measures were introduced to mitigate the increase (decontamination

campaign in 2005, increased injection of hydrogen and zinc), and at present, the

recirculation loop radiation fields seem to be low and well controlled.

− The annual occupational exposures at CNC have displayed a slightly increasing

trend during the last 10 years. This trend is explained by the combined effect of

increased radiation fields and considerable efforts spent in modifications and

maintenance during recent outages. The CNC annual exposures show large simi-

larities with the exposures in US BWRs of a similar design, but the US plants

generally show a slightly decreasing trend during recent years. A future decreas-

ing trend is also expected in the CNC plant due to the improved control of radia-

tion fields around the recirculation loops mentioned above.

− The general feeling is that the CNC power uprates have had a negligible impact

on occupational exposures, or at least are shadowed by more important factors

such as:

- water chemistry (HWC, copper, zinc, feedwater iron, etc.)

- fuel failures, recent failures have been caused by debris fretting, and do not

seem to be attributed to the power uprates

- some additional maintenance work, e.g. earlier postponed maintenance on the

RWCU system and experienced corrosion problems on the CRD piping in the

vessel pedestal penetrations.

Most of the works related to power uprates have concerned the turbine plant with

very low exposures. The only high dose rate work that could be attributed to the

power uprate (with some reservation) is the underwater repair of the steam dryer

during the 2005 refuelling outage. However, that work contributed rather little to

the total exposure during the outage.

78

4.3 Asco and Tihange

4.3.1 Introduction

Based on the data collected in the task #1, two PWR plants were selected for additional

analyses. Main criteria for selection were:

− Reactor design of the Westinghouse or Framatom type.

− Significant power increase, up to 10%.

− Steam Generator (SG) Replacement (New SG with tubes of Inconel 690 (i.e. a

high fraction of Ni in corrosion products).

− ≤18 months fuel cycles, boron – lithium chemistry.

− Plants where two years operational experience exists with the uprated plant, to

be able to assess the effects on the occupational doses.

Several PWR plants were considered but finally two PWR plants were selected for de-

tailed analyses: Asco and Tihange. NPP Asco and NPP Tihange are, by design, very simi-

lar plants. The fundamental operating basics at this type of Pressurised Water Reactor

(PWR) are set below.

Figure 4.3.1: A PWR plant – Operating diagram

79

The containment building (dry, ambient pressure, Figure 4.3.1) is the dominant and high-

est building of the plant designed to accommodate normal operating loads, functional

loads resulting from a loss-of-coolant accident, and the most severe loading predicted for

seismic activity. It provides biological shielding for both normal and accident conditions

and collection and holdup for leakage from the containment vessel. Inside the contain-

ment structure, the reactor and other NSSS components are shielded with concrete. The

primary circuit (Figure 4.3.2) consists of three independent loops connected to the reac-

tor vessel. Each loop includes a steam generator and a main pump. All these elements are

connected by a main pipe, forming a closed and completely watertight assembly to ensure

that all radioactive fluids are confined to the system. To ensure the continuous flow char-

acteristics and to prevent evaporation of the coolant water in the reactor, the entire system

is adequately pressurized relative to saturation pressure.

The power supply system to the reactor coolant pumps is designed so that adequate cool-

ant flow is maintained to cool the reactor core under all conceivable circumstances. All

pump parts in contact with the coolant are made of austenitic steel or stainless steel cov-

ered.

The steam generators, one per loop, are vertical U-tube units that contain Inconel 690

tubes.

The reactor coolant piping and all of the pressure containing and heat transfer surfaces in

contact with reactor water are stainless steel except the steam generator tubes and fuel

tubes, which are Inconel 690 and zircaloy or ZIRLO, respectively. Reactor core internals,

including control rod drive shafts, are primarily stainless steel.

An electrically heated pressurizer connected to one reactor coolant loop maintains RCS

pressure during normal operation, limits pressure variations during plant load transients,

and keeps system pressure within design limits during abnormal conditions. The pressur-

izer is located in one of the loops.

Figure 4.3.2: A PWR plant – Primary circuit

80

The turbine plant comprises a single turbine-generator unit. Saturated steam is supplied

from the steam generators to the turbine. The turbine is a 1500-rpm, tandem-compound

unit with one high pressure and two low-pressure cylinders. The steam flows through the

high-pressure turbine where it expands to low-pressure condition and then through mois-

ture separator reheater units where it is dried and reheated to superheated condition and

directed to the low-pressure turbines where it expands to condenser pressure condition.

The steam cycle is closed by condensing the steam from LP turbines, heating the conden-

sate in the feedwater system and returning the water to the steam generators through the

feedwater pumps.

The reactor pressure vessel (Figure 4.3.3) is cylindrical with a welded hemispherical

bottom head and a removable, bolted flanged and gasketed, hemispherical upper head.

The vessel contains the core, core support structures, control rods, and other parts directly

associated with the core. Outlet nozzles are arranged on the vessel to facilitate optimum

layout of the Reactor Coolant System equipment. The inlet nozzles are tapered from the

coolant loop vessel interfaces to the vessel wall to reduce loop pressure drop. The bottom

head of the vessel contains penetration nozzles for connection and entry of the nuclear in-

core instrumentation. Internal surfaces of the vessel that are in contact with primary cool-

ant, are weld overlay with stainless steel or Inconel.

Figure 4.3.3:A PWR plant – Reactor Vessel

81

The reactor core is composed of 157 fuel assemblies. The fuel assemblies are of square

cross section and are arranged in the reactor vessel in a way to optimally use the available

cylindrical space. Square spacer grid assemblies and the upper and lower end fitting as-

semblies support the fuel rods in fuel assemblies. Each fuel assembly is composed of 17 x

17 = 289 rods; of these, most of the places are used by fuel rods; the remaining places,

which are evenly and symmetrically distributed across the cross section of the assembly,

are provided with thimble tubes which may be reserved for control rods, and control in-

strumentation tube for incore thimble.

Control rod assemblies are used for reactor control and consist of clusters of stainless

steel clad neutron absorber rods inserted into guide tubes, one per tube. The absorber

rods move within the guide tubes. Above the core, each cluster of absorber rods is at-

tached to a spider connector and drive shaft, which is raised and lowered by a drive

mechanism mounted on the reactor vessel head. Downward stroke of the control rod

cluster is by gravity.

Water Chemistry

The RCS water chemistry is selected to minimize corrosion. A periodic analysis of the

coolant chemical composition is performed to verify that the reactor coolant quality meets

the required specifications. The Chemical and Volume Control System provide a means

for adding chemicals to the RCS which control the pH of the coolant during initial startup

and subsequent operation, scavenge oxygen from the coolant during startup and control

the oxygen level of the coolant due to radiolysis during all power operations subsequent

to startup. The pH control chemical employed is Lithium 7 hydroxide. This chemical is

chosen for its compatibility with the materials and water chemistry of borated wa-

ter/stainless steel/zirconium/inconel systems. In addition, lithium is produced in solution

from the neutron irradiation of the dissolved boron in the coolant. The concentration of

Lithium 7 hydroxide in the RCS is maintained in the range specified for pH control. If

the concentration exceeds this range, the cation bed demineralizer is employed in the

letdown line in series operation with the mixed bed demineralizer.

Dissolved hydrogen is employed to control and scavenge oxygen produced due to radio-

lysis of water in the core region.

Components with stainless steel sensitized in the manner expected during component

fabrication and installation will operate satisfactorily under normal plant chemistry condi-

tions in PWR systems because chlorides, fluorides and oxygen are controlled to very low

levels.

4.3.2 Asco NPP power uprate


Ascó NPP is located on the right bank of the Ebre River in the Ribera d'Ebre region. It is

located between the villages of Flix and Ascó (Tarragona), some 65-Km from the city of

Lleida and some 110-Km from the Ebre River’s mouth. The plant takes its name from the

82

village of Ascó, which is located 2-Km south of the plant. The village of Flix is located

north of Ascó NPP, holding the oldest chemical facilities in Spain.

Asco NPP consists of two 1028 MWe units, Westinghouse design 3 loop plant with 2912

Mwt pressurised water reactors. Supplier of the nuclear scope was Westinhouse and the

supplier of the non-nuclear design was Initec/Inypsa. Endesa have total ownership of

Asco 1 whereas at Asco 2, Endesa own 85% and Iberdorola 15%. Asco 1 had first criti-

cality June 17, 1983 and stared commercial operation December 10, 1984. Asco 2 had

first criticality September 11, 1985 and stared commercial operation March 31, 1986.

Both units have been uprated since commissioning. The thermal power of Asco 1 reactor

was increased from 2696 MW to 2900 MW in 2000 and to 2951 MW in 2003. The corre-

sponding nominal values of the net electrical output were 930 MWe, 1028 MWe and

1033 MWe, respectively. The present study focuses on the Asco 1 first uprate, resulting

in a thermal power level of 8% compared to the initial power level. The net electrical

output from the plant during the period 1983 – 2005 is shown in Figure 4.3.4.

The latter uprate was an uprate of 1.5 %, achieved using more precise techniques for

measuring feedwater flow.

Capacity and Annual production

540560580600620640660680700720740760780800820840860880900920940960980

1000

1982 1984 1986 1988 1990 1992 1994 1996 1998 2000 2002 2004

Years

MW

e

Safety

analyses

upgradeSGRcapacity

annual production

Improvement of FW

measurement

Figure 4.3.4: Asco 1 – Net electrical output

The reactor core of Asco 1 contains of 157 fuel assemblies, which means 18.80 MWth

per fuel assembly at present maximum power level, compared to 17.20 MWth per assem-

bly at initial design power. The fuel assemblies are 17x17 design, for the first cores

STANDARD design, with average enrichment 3.5 %. Since then new fuel designs have

been adopted. At present it is MAEF-IFM+RP (Enusa) with average enrichment 4.7%.

Each 17x17 MAEF+IFM+RP is composed of a bundle of 264 fuel rods. These rods are

arranged in a square lattice of 17 x 17 positions and are supported by twelve grids, two of

which are end grids, six intermediate, three mixer and one protective. The grids, together

83

with 24 thimbles tubes, an instrumentation tube and two nozzles at the ends, form the

structural skeleton of the fuel tube. Main specification of fuel is given below.

84

A summary of the modifications that have been introduced in the Asco 1 plant during the

period 1995 – 2003 is presented in Table 4.3.1. Not all modifications are due to the

power uprates, but in many cases form part of the ongoing plant modernization programs

that have been carried out. Those two aspects are not always easy to separate; introduced

modifications are in many cases addressing both aspects. The major modernization work

was performed in 1995-1997 when was replaced SG and both HP and LP turbines were

modified. Electrical power was increased after turbine modifications (1995, 1997), ther-

mal power was increased in year 2000 when upgrade of Safety Analyses was performed.

Table 4.3.1 – Asco 1 Summary of modernization work

Year Work performed

1995 SG Replacement

Modification of High Pressure Turbine

1997 Modification of Low Pressure Turbine

2000 Upgrade of Safety and System Analysis

Turbine modification for increased flow

Upgrades of main electric equipment

2003 Improvement of the feed water flow measurement

The variation in outage lengths in period 1990 – 2004 is presented in Figure 4.3.5. On the

beginning of operation outages duration were mostly about 40 days. Planned outages

show trend of shortening during last decade. In the year 1995 outage was prolonged due

to SG replacement and RTD Bypass replacement. Impact on occupational exposures is

addressed later in the report.

Outage lengths

0

10

20

30

40

50

60

70

80

90

100

1990 1991 1992 1993 1994 1995 1996 1997 1998 1999 2000 2001 2002 2003 2004

years

days

planned outage

forced outage

Figure 4.3.5: Asco 1 – Outage lengths for years 1990-2004

85


Detailed analyses of chemistry and radiochemistry were not possible to perform due to

non-availability of that data. However, a short Reactor Coolant System (RCS) chemistry

specifications were available:

Parameter Value

Solution pH Determined by the concentration of boric acid and

alkali present. Expected values range between 7.04

(high boric acid concentration) to 7.1 (low boric

acid concentration)

pH Control Agent (Li7OH) 3.5 ppm

Boric Acid, ppm B Variable from 0 to approximately 3000



A way to assess the overall affect of power uprates on radiation levels during operation is

to study the occupational exposures obtained during reactor operation. Data from Asco 1

are summarized in Figure 4.3.6. It must be noted, that the annual exposure during opera-

tion is normally a small fraction, 2-20%, but mostly less than 10%, of the total occupa-

tional exposure. Therefore, the data in Figure 4.3.6 shows some variation, but no clear

correlation to the power uprate is seen.

Annual ocupational exposure - normal operation

0

50

100

150

200

250

300

350

400

450

500

550

600

1990 1991 1992 1993 1994 1995 1996 1997 1998 1999 2000 2001 2002 2003years

Exp

osu

re (

man

mS

v)

Figure 4.3.6: Asco 1 – Annual occupational exposure

86


The dose rates during outage conditions have been regularly followed-up. Measured dose

rates on survey points around primary piping (primary circuit and channel head hot and

cold legs) at outage conditions are shown on Figure 4.3.7. The dose around the channel

head is the biggest fraction, approximately 95% of the total measured dose on survey

points. From the figure it is obvious that radiation level decreases, starting with SG Re-

placement, year 1995.

Dose rates on survey points **

0306090

120150180210240270300330

1990 1991 1992 1993 1994 1995 1996 1997 1998 1999 2000 2001 2002 2003

years

Do

se R

ate

(m

Sv/h

)

Cold Leg Hot Leg

Figure 4.3.7: Asco 1– Dose rates on survey points**

**survey points: Channel head /Primary circuit


The annual occupational exposures during the period 1990 – 2003 are presented in Fig-

ure 4.3.8. The exposures are divided into contribution during outage and operation condi-

tions. The exposures during operation are only a small fraction of the total exposures, and

have trend shown on the Figure 4.3.6. A clear correlation with the outage lengths (Figure

4.3.5) is seen, i.e. outages involving large efforts for the power uprate and plant moderni-

zation projects have consequently implied increased exposure).

Data on exposure per outage day are compiled in Figure 4.3.9. The exposure per outage

day shows some variation, but no trend of increasing exposure after SG Replacement and

the power uprate is seen.

87

Annual ocupational exposures

0

500

1000

1500

2000

2500

3000

3500

4000

4500

5000

5500

6000

1990 1991 1992 1993 1994 1995 1996 1997 1998 1999 2000 2001 2002 2003years

Exp

osu

re (

man

mS

v)

operation

outage

Figure 4.3.8: Asco 1 Annual occupational exposures split-up on outage and operation

conditions

Occupational exposure per outage day

0

20

40

60

80

100

120

140

160

180

1990 1991 1992 1993 1994 1995 1996 1997 1998 1999 2000 2001 2002 2003 2004

years

Exp

osu

re (

man

mS

v)

Figure 4.3.9: Occupational exposures per outage day

The Asco 1 annual exposures are compared to some international PWR data in Figure

4.3.10. The Asco 1 exposures are in the last decades rather low compared to the average

international PWR. Only years with considerable efforts in modernization projects, e.g.

1995-1997, imply higher exposures comparable with the average European plants.

88

Figure 4.3.10: Average outage collective dose per reactor type and per country


A review of data and experience from the Asco 1 plant has been performed. The review

has resulted in the following conclusions:

− The plant has been uprated twice since the commissioning. The thermal power of

reactor was increased from 2696 MW to 2900 MW in 2000 and to 2951 MW in

2003. The SG Replacement was performed in 1995 but all safety analyses neces-

sary for power uprate were performed in 2000. The present study has focused on

the first uprate, resulting in a thermal power level of 8% compared to the initial

power level. The latter uprate was an uprate of 1.5 %, achieved using more pre-

cise techniques for measuring feedwater flow.

− In year 1995 when a SG Replacement was performed the planned outage length

was 34 days. The reality was it took 61 forced outage days. Typical outage

lengths are between 30 and 40 days.

− Standard PWR water chemistry was maintained, no zinc is injected.

− Operational exposure has been maintained on a constant and rather low level

since the uprate.

− Dose rates, during outage, on reactor systems have been maintained on a rather

low and constant level after SG Replacement.

− Considerable manhour efforts spent during some outages for the plant moderniza-

tion program have, of course, resulted in some increase of occupational expo-

sures. The exposures per outage day have decreased in the last decade. The aver-

age annual exposures in the Asco 1 have been kept on a level comparable to

international values for PWR plants.

89

4.3.3 Tihange NPP power uprate


Tihange NPP is located on the bank of the Meuse River in the Liege region. It is located

in Huy, some 30-Km from the city of Liege, the municipality includes the old commune-

sof Ben-Ahin and Tihange.

Tihange NPP consists of three units, Framatome/Westinghouse design 3 loop plant with

pressurised water reactors. Westinghouse and Framatome, associated with Alstom ACEC

Energie and Cockerill Mechanical Industry (CMI) supplied the nuclear scope, the sup-

plier of the non nuclear design was Alsthom ACEC. Tihange 1 is 50% owned by Elec-

trabel 50% and 50% by EDF. Whereas Electrabel own 96% of Tihange 2 & Tihange 3

and the Belgian utility SPE own the remaining 4%. Tihange 1 had first criticality Febru-

ary 21, 1975 and started commercial operation on October 1, 1975. Tihange 2 had first

criticality October 13, 1982 and started commercial operation on Jun 6, 1983. Tihange 3

had first criticality Jun 5, 1985 and started commercial operation on September 1, 1985.

Tihange 1 and Tihange 2 have been uprated since the commissioning. The thermal power

of Tihange 1 reactor was increased from 2665 MW to 2875 MW in 1995. The corre-

sponding nominal values of the net electrical output were 914 MWe, 1009 MWe. The

thermal power of Tihange 2 reactor was increased from 2785 MW to 2905 MW in 1995

and to 3064 MW in 2001. The corresponding nominal values of the net electrical output

were 945 MWe, 1007 MWe and 1055 MWe. The present study focuses on the Tihange 2

second uprate, resulting in a thermal power level of 10% compared to the initial power

level. The net electrical output from the plant during the period 1983 – 2005 is shown in

Figure 4.3.11.

Capacity and Annual production

600620640660680700720740760780800820840860880900920940960980

10001020

1983 1985 1987 1989 1991 1993 1995 1997 1999 2001 2003 2005

Years

MW

e

SG ReplacementMOX fuelcapacity

annual production

Figure 4.3.11: Tihange 2 – Net annual electrical output and Capacity

90

The reactor core of Tihange 2 contains 157 fuel assemblies, which means 19.52 MWth

per fuel assembly at present maximum power level, compared to 17.74 MWth per assem-

bly at initial design power. The fuel assemblies are 17x17 design, for the first cores

STANDARD design, with enrichments 1.6%/2.2%/3.5 %. Since then, new fuel designs

have been adopted, MOX (Belgonucleaire) with average enrichment 3.8%. Each fuel

assembly is composed of a bundle of 264 fuel rods. A MOX element is simply a fissile

element that weighs approximately 450 kg and is composed of a mixture of 415 kg of

uranium oxide and 35 kg of plutonium oxide. By replacing a standard fuel element com-

posed of enriched uranium by a MOX element, 9 kg of plutonium is consumed instead of

5 kg that is produced.

The recovering of plutonium in MOX elements reduces the quantity of plutonium pro-

duced in nuclear power plants and economizes uranium ore needs.

A summary of modifications that have been introduced in the Tihange 2 plant during the

period 1995 – 2001 is presented in Table 2. Modifications are not mainly due to the

power uprates, but in many cases are part of ongoing plant modernization programs.

Those two aspects are not always easy to separate; introduced modifications are in many

cases addressing both aspects. Major modernization work was performed in 1995 when a

new fuel design was introduced, and in 2001 when SG were replaced and a revision of the

Safety Analyses was performed. In both cases thermal power was increased.

Table 4.3.2 – Tihange 2 Summary of modernization work

Year Work performed

1995 First loading of MOX fuel in Tihange 2 (March)

2001 During the outage of Tihange 2, which started

on 9 June and ended on 11 August, all three

Steam Generators were successfully replaced.

The steam generator replacement was executed

in the new record time of 17.5 days.

Revision of safety analyses

The variation in outage lengths 1990 – 2004 is presented in Figure 4.3.12. Normal plant

outages duration is about 30 days. In 1997 the outage was prolonged due to RCP inspec-

tions and repairs. In 2001 the outage was prolonged due to SG replacement. Plant avail-

ability is mostly about 86%, and the modernization and power uprate programs have not

impacted on the average outage length (impact on occupational exposures is addressed

later in the report).

91

Outage length

05

1015

2025

3035

4045

5055

6065

1990 1991 1992 1993 1994 1995 1996 1997 1998 1999 2000 2001 2002 2003 2004

years

days

Figure 4.3.12: Tihange 2 – Outage lengths for years 1990-2004


Detailed analyses of chemistry and radiochemistry were not possible to perform due to

non-availability of that data. However, a short Reactor Coolant System (RCS) chemistry

specifications were available:

Parameter Value

Solution pH Determined by the concentration of boric acid and

alkali present. Expected values approximately 7

pH Control Agent (Li7OH) 3.5 ppm

Boric Acid, ppm B Variable from 0 to approximately 3000



A way to assess the overall affect of power uprates on radiation levels during operation is

to study the occupational exposures obtained during reactor operation. Data from Tihange

2 is summarized in Figure 4.3.13. It must be noted, that the annual exposure during op-

eration is normally only a small fraction, 5-20% of the total occupational exposure.

Therefore, the data in Figure 4.3.13 shows some variation, but no clear correlation to the

power uprate is seen. The occupational exposure has been decreasing in the last decade.

92

Annual ocupational exposure - normal operation

0

50

100

150

200

250

300

350

400

450

500

1990 1991 1992 1993 1994 1995 1996 1997 1998 1999 2000 2001 2002 2003 2004

years

Ex

po

su

re (

ma

n m

Sv

)

Fi

gure 4.3.13: Tihange 2 – Annual occupational exposure

4.3.3.3.2 During outage

Dose rate during outage conditions have been regularly followed-up. Measured dose rates

on survey points around primary piping (primary circuit and channel head hot and cold

legs) at outage conditions are shown on Figure 4.3.14. The dose around the channel head

is the biggest fraction, approximately 95%, of the total measured dose on survey points. It

can be seen that radiation level decreases after SG replacement, year 2001.

93

Dose rates on survey points**

0

50

100

150

200

250

300

1990 1991 1992 1993 1994 1995 1996 1997 1998 2000 2001 2002 2003

years

Do

se r

ate

(m

Sv/h

)

cold leg hot leg

Figure 4.3.14: Tihange 2 – Dose rates on survey points**

* There was an outage in the year 2002 but dose rate data is missing for

year 2002

**survey points: Channel head / Primary circuit


The annual occupational exposures during the period 1990 – 2003 are presented in the

Figure 4.3.15. The exposures are divided into contribution during outage and operation

conditions. The exposures during operation are only a small fraction of the total expo-

sures; the trend is shown on the Figure 4.3.13. Data on exposure per outage day is com-

piled in Figure 4.3.16. The exposure per outage day shows some variation, but what is

important, it that it a shows decreasing trend over the last decade. The outages are respon-

sible for the major part of the collective doses. In the year 2001, the steam generator re-

placement was responsible for half of the collective dose in Tihange 2.

There is a correlation with the outage lengths (Figure 4.3.12), i.e. outages involving large

efforts for the power uprate and plant modernization projects have consequently implied

increased exposure. It can also be seen from exposure data for 1997, that the outage

length was longer than most of the others but the exposure was not very high; no bigger

modernization activities were performed.

94

Annual ocupational exposures

0

500

1000

1500

2000

2500

3000

1990 1991 1992 1993 1994 1995 1996 1997 1998 1999 2000 2001 2002 2003 2004

years

Ex

po

su

re (

ma

n m

Sv

)

operation

outage

Figure 4.3.15: Tihange 2 Annual occupational exposures split-up on outage and opera-

tion conditions

Ocupational exposure per outage day

0

10

20

30

40

50

60

1990 1991 1992 1993 1994 1995 1996 1997 1998 1999 2000 2001 2002 2003 2004

years

Exp

osu

re (

man

mS

v)

Figure 4.3.16: Tihange –Occupational exposures per outage day

The Tihange 2 annual exposures are compared to Belgium and some international data in

Figure 4.3.17. Collective doses in Tihange are increasing compared to 2000. This is due

to the outage and steam generator replacement. In 2003 and 2002 there was outage in

Tihange 2. In 2003, there was a supplementary stop at Tihange 2 for pressuriser’s leg

welding inspection. There was insufficient data after SG replacement to evaluate the real

exposure trend but it seems to be decreasing.

95

Figure 4.3.17: Average outage collective dose per reactor type and per country


A review of data and experience from the Tihange 2 plant has been performed. The re-

view has resulted in the following conclusions:

− The plant has been uprated twice since the commissioning. The thermal power of

the reactor was increased from 2785 MW to 2905 MW in 1995 and to 3064 MW

in 2001. The present study focuses on the Tihange 2 second uprate, resulting in a

thermal power level of 10% compared to the initial power level.

− In the 2001 three SG were replaced in 63 days. Typical outage lengths are be-

tween 30 and 40 days. The total annual exposure in 2001 was 1539 man mSv.

Outage exposure was 1446 man mSv, and half of it was due to SG replacements,

which indicates good organization.

− Standard PWR water chemistry is maintained, no zinc is injected.

− Since the uprate exposures during operation have been maintained on a constant

and rather low level.

− Radiation levels during outage on reactor systems have been maintained on a

rather low and constant level after SG Replacement.

− Considerable manhour efforts spent during some outages for the plant moderniza-

tion program have of course resulted in some increase of occupational exposures.

The exposures per outage day have shown a decreasing trend in the last decade.

The average annual exposures in the Tihange 2 have been kept on a level compa-

rable to international values for PWR plants.

Tihange 2

96

4.3.4 Comparison of two PWR uprates

For two selected PWR plants, Asco 1 and Tihange 2, the SG replacement and the mod-

ernisation activites were analysed for tehri contribution to occupational doses.

Typical values for:

− Outages length

− Number of personnel (plant and outside)

− Collective dose by job and type of personnel

− Collective dose by task and type of personnel

− Collective dose by occupational category and type of personnel

were reviewed and are presented in the Table 4.3.3 and Figure 4.3.18-Figure 4.3.25.

Table 4.3.3: Specific elements of the PWR plants for the year of uprate

ASCO 1 TIHANGE 2

Annual collective dose by type

of personnel

(man mSv)

• Plant

• Outside

129

5239

152

1387

Outage length

(SGR + other outage activities)

95 days

61 days per SGR

63 days

52.5 days per SGR

Number of outage personnel

• Plant

• Outside

335

2057

324

1235

Collective outage dose

(man mSv)

• Plant

• Outside

Planned out. Forced out.*

* SGR

113 1.35

1329 3827

Planned out. Forced out.

90 -

1356 -

Collective dose by task

(man mSv)

SG Replacement

2443 man mSv

648 man mSv

97

If normal tasks are compared with tasks performed in the outage per particular year:

− Refuelling

− Reactor Vessel activities (Inspection, Maintenance)

− Activities on SG Primary side

− Activities on SG Secondary side

− HRS&SIS systems

− RC pumps

− Valve work

− Routine Inspection

− Scaffolding

− Insulation

− Large tasks (as SG Replacement)

There is no big difference in the extent of work per outage, there are some deviations but

not significant. Most of the activities at Tihange plant are performed with smaller occupa-

tional exposure. Table 4.3.3, shows that a large task like SG Replacement is performed in

a more optimised manner at Tihange plant.

If a comparison of exposure per occupational category in both plants is made, high expo-

sure is received by mechanical personnel, inspection personnel and decontamination per-

sonnel

The following tables show comparisons for Tihange and Asco for the selected task of SG

replacement. The comparison period was two years prior to the replacement and two

years after. It has to be noted that Asco 1 performed SG replacement in 1995 and Tihange

2 in 2001.

98

Total annual exposure

0500

10001500200025003000350040004500500055006000

two years before year before SGR year year after two years after

Exp

osu

re (

man

mS

v)

Asco Tihange

Figure 4.3.18: Total annual exposure for five years period

** Asco 1 did not have outage in the year after SG Replacement

Tihange 2 did not have outage two years before SG Replacement

Normal operation exposure

050

100150200250300350400450500550600


Exp

osu

re (

man

mS

v)

Asco Tihange

Figure 4.3.19: Normal operation exposure for five years period

99

Outage exposure

0500

10001500

20002500

30003500

40004500

50005500

6000


Exp

osu

re (

man

mS

v)

Asco Tihange

Figure 4.3.20: Outage exposure for five years period



Outage length

0

20

40

60

80

100


days

Asco Tihange

Figure 4.3.21: Outage length for five years period

100



Normal operation - plant personnel

0

20

40

60

80

100


Exp

osu

re (

man

mS

v)

Asco Tihange

Figure 4.3.22: Normal operation exposure - plant personnel

Normal operation - outside personnel

0

50

100

150

200

250

300

350

400

450

500


Exp

osu

re (

man

mS

v)

Asco Tihange

Figure 4.3.23: Normal operation exposure-outside personnel

101

Outage operation - plant personnel

0

20

40

60

80

100

120

140

160

180

200


Exp

osu

re (

man

mS

v)

Asco Tihange

Figure 4.3.24: Outage operation exposure-plant personnel



Outage operation - outside personnel

0

5001000

15002000

25003000

35004000

45005000

5500

two years

before

year before SGR year year after two years after

Exp

osu

re (

man

mS

v)

Asco Tihange

Figure 4.3.25: Outage operation exposure-outside personnel

102



From the figures 4.3.18-4.3.25 and Table 4.3.3 following conclusions emerge:

• For the five year period (two years before and after SGR), total exposure, including

outage and normal operation exposures are greater for Asco plant. The only exception

is the year when there was no outage in the Asco plant. Asco plant has longer outages

than Tihange plant and contractors are more loaded. Even during normal operation

most of the activity in the Asco plant is performed by outside personnel.

• There are some visible differences in exposures data for Asco 1 and Tihange 2 unit.

Some facts have to be underlined:

− Leadership, composition and organisation of the power uprate, including radio-

logical activities, especially large demanding tasks are critical for successful im-

plementation of power uprate and received doses. When exposures are, compared

for SG Replacement it can be noted that occupational exposure in Asco 1 was

almost four times higher than in the Tihange 2. However, it must be noted that

SG replacement in Asco was performed in year 1995, and SG Replacement in

Tihange was performed in year 2001.

− A prerequisite for applying the principle of optimisation to occupational radiation

protection is the appropriate and timely exchange of data, techniques and experi-

ence on doses and dose reduction methods. The Information System on Occupa-

tional Exposure (ISOE) was launched in year 1992 after a two-year pilot pro-

gramme, with aim of helping utilities in sharing information and experience

worldwide. Thus, Tihange plant had more available data,..

− Most of the activity in normal operation and outage in the Asco plant are per-

formed by outside personnel. Event reports from worldwide Operating Experi-

ence demonstrate that it is not recommended to allocate the responsibility for im-

portant activities to outside personnel without strong supervision by plant

personnel. Outside personnel are not as well-trained or as acquainted with plant

design details as plant personnel. If outside personnel perform activities they

have to be trained and retrained to stress the potential for abnormally high or rap-

idly changing radiological conditions, including the the actions required when

these conditions occur. Training should also emphasize the importance of a high

level of awareness and sense of individual responsibility with regard to personnel

radiation protection.

• Although the exposures are higher for Asco plant, what should be emphasised is the

decreasing trend of exposures during last decade. The most important impact in the

reduction of collective dose was the removal of the old RTD System, the SG Re-

placement, the substitution of the satellite and the ALARA Program.

103

5. Reconstruction experience

The aim of task#3 was to identify good (and bad) practices that are impacting on the

doses to personnel and could be related to an uprate. From the perspective of occupational

exposure it should give an answer to what kind of design, implementation and operational

arrangement are the best, Optimisation of the work processes to limit the duration of the

time spent in the controlled areas is specially highlighted. Leadership, composition and

organization of the large demanding tasks are critical for successful implementation of

power uprate and received doses

5.1 Technical factors

5.1.1 Introduction

In the following sections, analysis of the outputs of task 2 and task 3 of the project have

consisted of a review of the important factors in BWRs and PWRs which affect radiation

levels and occupational exposures in general, and especially at power uprates.

The following sections review technical factors important for radiation fields in BWRs

and PWRs, and how these are affected by power uprates.

5.1.2 BWR uprates

5.1.2.1 Water chemistry issues

5.1.2.1.1 Corrosion product balance

The water chemistry control in BWRs to combat radiation buildup is largely based around

the optimisation of the corrosion product balance in the primary circuit. Six different

general types of corrosion product balance are schematically illustrated in the Figure

5.1.1. It is concluded that the fuel crud composition should be well balanced, i.e. the rela-

tion between iron (Fe) and nickel (Ni) plus zinc (Zn) should be maintained close to the

spinel relation, with only a small excess of Fe. A significant inflow of Fe may results in

fuel corrosion problems, especially if the Fe inflow is occurring together with a consider-

able inflow of Zn (and copper6 (Cu)). High inflow of Fe normally results in a fuel crud

that is prone to release particles resulting in hot spot radiation sources in the plant. How-

ever, a too low inflow of Fe may lead to formation of less stable monoxides, resulting in

increased reactor water activity concentrations, and, particularly in connection to some-

what increased steam moisture content and increased turbine plant radiation levels. Fuel

crud with a too high Ni/Fe ratio also seem to be involved in increased local fuel cladding

corrosion, especially for fuel with spacers of Inconel.

6 Cu shall be avoided for several reasons. Cu is known in several cases to cause fuel cladding corrosion

(CILC). Cu also makes HWC operation less effective. There are, however, some indications that a moder-

ate amount of Cu in the fuel crud may form oxide forms that have a high affinity for cobalt (Co), resulting

in reduced Co-60 reactor water activity.

104

The above general recommendations indicate different actions whether the plant is oper-

ating with NWC, HWC or has applied NMCA.

Ni/Fe > 0.5

Zn

(Zn+Ni)/Fe << 0.5

Ni/Fe << 0.5

Zn(Zn+Ni)/Fe > 0.5

Ni/Fe ≤ 0.5NWC

Zn(Zn+Ni)/Fe ≤ 0.5

HWC, NMCA

?

Fuel problems

Turbine activity

Particle problems

Fuel problems

Figure 5.1.1: BWR Corrosion product balance Fe/Ni/Zn - Six different cases: Where to go?

The Fe inflow in plants operating with NWC is very much dominated by the inflow by

the feedwater. That means that the key to controlling the fuel crud composition is to con-

trol the feedwater chemistry, and especially the feedwater Fe. The most recent EPRI wa-

ter chemistry guideline proposes that the feedwater Fe concentration during NWC condi-

tions should be maintained in the interval 0.5 – 1.5 ppb. The proposed amount of Fe is

probably rather conservative, especially in the case with low feedwater Zn. Experience

from Scandinavian BWRs has shown, that a well balanced fuel crud with respect to Fe

and Ni can be maintained with a feedwater Fe concentration as low as about 0.2 ppb, if

the Zn level is low. This level is supported by typical fuel crud Ni amounts measured. Zn

in the feedwater, natural or injected, increases the amount of feedwater Fe needed to

maintain a fuel crud of the spinel type:

Eq. 1 0.2 2.3FW FW

Fe ZnC C≥ + ⋅

where: FW

CFe – Feedwater Fe concentration [ppb] FW

CZn – Feedwater Zn concentration [ppb]

A control of the feedwater Fe according to Eq. 1 will result in a fuel crud close to the

ideal spinel type, i.e. further injection of Zn is not needed and will only result in an in-

creased demand of feedwater Fe. On the other hand, if the minimum feedwater Fe is not

easily obtained, Zn injection can help to improve the characteristic of the fuel crud. The

1.5 ppb EPRI 2004 upper bound of feedwater Fe corresponds to a feedwater Zn level of

maximum 0.6 ppb, which corresponds to the recommended maximum EPRI level. Higher

Zn (and Fe) levels may result in increased fuel cladding corrosion.

105

The prerequisites for Fe control change considerably when HWC operation is applied,

and even more when NMCA application is performed. The feedwater is no longer the

only source of Fe, considerable contribution is also expected from sources in the reactor

circuit obtaining low corrosion potential. These internal sources are not easily monitored,

and significant variation may exist between different plants. Other than reactor design

features, the influencing factors are the degree of H2 injection in the case of HWC, the

degree of NM coverage in the case of NMCA, and the pre-history with respect to feedwa-

ter Fe and Zn inflow. The recent EPRI recommendations consider this effect, and low

feedwater Fe, acceptable. Recommended interval for feedwater Fe is 0.1 – 1 ppb. The

lower limit represents the limit of today’s US BWR experience, and a practical interpreta-

tion is that actually no lower limit exists in the case of HWC and NMCA plants. In prac-

tice, large efforts are made in US plants to lower the feedwater Fe input.

In the case of HWC and NMCA plants, Zn injection seems to be especially effective in

forming a more stable fuel crud composition. However, the amount Zn injection needed is

not so easily determined due to the above mentioned, non-monitored internal sources of

Fe. The recent EPRI guidelines propose that a reactor water Zn level of >5 ppb shall be

maintained in HWC plants. In the case of NMCA plants a relation between soluble 60

Co

and Zn is proposed instead (<2.0·10-5

µCi/g per ppb, or <720 Bq/kg per ppb), which in

reality, normally means a somewhat lower reactor water Zn level than 5 ppb. However,

the reactor water specifications have to consider the proposed feedwater Zn limits, <0.6

ppb in HWC plants and <0.4 ppb in NMCA plants, which may override the reactor water

limits. The feedwater Zn limits are due to fuel concerns.

As mentioned above, the Zn injection may be complicated to control due to the non-

monitored sources of Fe in HWC and NMCA plants. Therefore, one Scandinavian HWC

plant with low feedwater Fe has used an alternate way of controlling the feedwater Zn

injection based on relation between reactor and feedwater Zn:

Eq. 2 2FW RWRWCUZn Zn

FW

fC C

f≥ ⋅ ⋅

Eq. 3 RW Max

Zn ZnC C≤

where: FW

CZn – Feedwater Zn concentration [ppb] RW

CZn – Reactor water Zn concentration [ppb] Max

CZn – Maximum allowed reactor water Zn concentration [ppb]

fRWCU – Reactor water cleanup flow [kg/s]

fFW – Feedwater flow [kg/s]

The above relation Eq. 2 means that at least about half of feedwater Zn shall be consumed

in restructuring of fuel crud and system surface oxides, and maximum about 50% of the

feedwater Zn is allowed to be cleaned-up by the RWCU. The feedwater Zn is primarily

adjusted to reach the reactor water Zn target level, Max

CZn (Eq. 3). If the Zn target level

cannot be reached together with the relation Eq. 2, the Zn injection is decreased to a point

where Eq. 2 is fulfilled and a somewhat lower reactor water Zn level than the target is

106

accepted. This operation strategy is to ensure that a certain iron surplus in the fuel crud is

maintained.

A power uprate in a BWR plant certainly has the potential to affect the corrosion product

balance. The demands on the condensate polishing plant are normally increased due to

higher flow rates, increased operation temperature and higher inflow of Fe from erosion

corrosion. In many uprate projects the turbine design is also changed from a full-flow

condensate cleanup to forward pumped heater drains with less cleanup of the condensate.

In other words, there is a risk that the outcome of a power uprate is increased feedwater

inflow of corrosion products, especially Fe, which has to be addressed in the project. On

the other hand, the turbine design modifications included in a power uprate project imply

a potential to introduce improvements. Replacements of e.g. steam extraction pipes from

previous carbon steel to low-alloy steel will significantly reduce the erosion corrosion Fe,

something that is especially important in forward pumped designs (see Section 4.2). An-

other modification sometimes introduced is the revamping of the turbine condenser from

previous brass to titanium with a reduction of the Cu inflow to the primary circuit (see

Section 4.2). Furthermore, large efforts are spent in many BWRs to improve the perform-

ance of the condensate polishing plants. An example of such an improvement are the

modifications performed in the OL1/2 plants (see Section 4.1). A review among the 39

BWRs (36 North American, 3 European) that participated in the EPRI BWR chemistry

monitoring database in 2005 has been made and was reported [2] at the 2006 water chem-

istry conference in Korea. There are three configurations of condensate polishing systems

employed: deep bed demineralizers only, filter + deep bed, and precoated filter deminer-

alizers. Feedwater iron, feedwater copper, and reactor water sulphate are three key pa-

rameters that are of particular interest for assurance of fuel reliability, mitigation of inter-

granular stress corrosion cracking (IGSCC), and achievement of dose reduction goals.

Deep bed (DB) demineralizers have a high capacity to remove ionic impurities but lim-

ited capability to remove particulate impurities (such as iron and copper oxides).

Filter + Deep Bed (F+DB) systems have deep bed demineralizers downstream of high

efficiency backwashable filters. These filters are commonly called prefilters and are used

without precoat materials. In North America, all prefilter systems currently employ filter

septa with pleated filter media. The Filter + Deep Bed design provides both efficient re-

moval of insoluble crud particles and high capacity to remove ionic impurities.

Filter Demineralizers (FD) employ vertical cylindrical filter septa (either pleated or

wound) that are precoated with filter aid materials made up totally or partially of pow-

dered cation and anion exchange resins. This design provides efficient removal of insolu-

ble crud particles but has only limited capacity to remove ionic impurities.

Measures taken by the industry to improve iron control have been successful. This is evi-

dent in the downward trend from 1997-2005 in average feedwater iron results for 39

BWRs shown in Figure 5.1.2. A 63.8% reduction in average feedwater iron occurred

between 1997 and 2005, with a 19.3% reduction between 2004 and 2005, as shown in

Figure 5.1.3, this trend is expected to continue as more Deep Bed Only plants retrofit

prefilters and as more Filter Demineralizer plants implement the use of lower particle

retention rating septa.

107

Figure 5.1.2: BWR Feedwater Iron Trend from 1997-2005

for 39 BWRs by Condensate Polishing System [2]

Figure 5.1.3: Individual BWR Feedwater Iron Trend for 2004 and 2005 [2]

When planning a power uprate, the effects of increased flow and increased inlet iron must

be addressed. The effects of increased flow and increased inlet iron are multiplicative;

i.e., a 10% increase in flow combined with a 10% increase in inlet iron results in a 21%

increase in filter iron loading.

Increased condensate flow results according to [2] in:

− Higher area flow rates which affect filtration and ion exchange;

108

− Higher filter dP resulting in shorter runs and shorter septa life and higher rad-

waste volumes; and

− Lower system dP margin which may further limit useful septa life.

Increased inlet iron results in [2]:

− Higher dP resulting in shorter runs and higher radwaste volumes; and

− More solids to be processed in radwaste than would be expected by flow increase

alone.

Stations have addressed these issues in a variety of ways, including:

− Installation of an additional service vessel;

− Increasing septa length and/or diameter;

− Planned bypass flow during times when a vessel is out of service for backwash-

ing.

The overall BWR trend is improved feedwater chemistry in spite of power uprate projects

in many BWRs.

5.1.2.1.2 Cobalt sources

Co-60 is the dominant radiation source in BWR plants, and it is, therefore, of utmost im-

portance to assess, and possibly reduce, the sources of cobalt (Co) in the plant. The meas-

urement of the very low concentrations of Co in feed- and reactor water is not an easy

task, and the industry has spent lots of efforts to map the actual Co sources. It has become

evident, that the material that contributes most to the Co source is Stellite, which contains

about 60% of Co. This material is especially used in valve seats and in pumps, in both

reactor and turbine systems. Some low-pressure turbines also have Stellite on the blade

tips. Of special interest is the use of Stellite as “pins and rollers” on control rods. The

Stellite on control rods is neutron activated, and the corrosion means an inflow of Co-60

to the primary circuit in parallel to the Co inflow. Many plants have been successful in Co

reduction programs, but many applications with Stellite are not easily replaced with Co-

free substitutes.

Available Stellite replacement materials are further discussed in the PWR section.

Forsberg et al., 2004 [3], report laboratory measurements of Co release from Stellite in

different BWR water chemistries, see Figure 5.1.4. Release rates are lower in a reducing

environment than in a oxidising one and dosage with Zn causes a further reduction. A

change from NWC conditions without Zn to HWC conditions with Zn causes a reduction

in Co release by almost an order of magnitude. The effect is one of several explanations

for the beneficial effect of Zn injection, especially at HWC conditions.

109

Figure 5.1.4: Laboratory testing of Co release rate from Stellite –

Influence of HWC and Zn, Forsberg et al., 2004 [3]

Stellwag & Staudt, 2004, report data from a German plant indicating that the replacement

of control rods containing Stellite with Co-free substitutes has resulted in a significant

reduction of reactor water Co-60, see . 116 out of 145 control rods were replaced, and an

about 50% reduction of the reactor water Co-60 was observed, and a corresponding long-

term reduction of radiation fields are expected. Similar assessments have been made for

other BWRs, and somewhat lower contribution to the Co-60 activity has been obtained.

Figure 5.1.5 Reduction of Co-60 RW activity in a German BWR due to control rod re-

placement without Co containing buttons and rollers (116 out of 145 CRs replaced)

(green line: Co-60, red line: Co-58, blue line: Zn-65 [Bq/Mg]), Stellwag & Staudt, 2004

110

Plants with full-flow condensate cleanup normally have only a minor contribution to the

total Co inflow from the turbine plant, i.e. the major Co contribution is from Stellite in

reactor systems. Plants with turbine design employing forward pumped heater drains

(FPHD) normally display feedwater Co input that is comparable or even higher than the

reactor system sources, i.e. FPHD plants normally have a somewhat higher total input of

Co to the primary circuit than full-flow cleaned-up plants.

Most plants have Co reduction programs, where a gradual replacement of Stellite in

valves and other applications are planned. However, a problem with such programs is the

lack of knowledge of the exact Co contribution from different sources7. The outcome of

performed modifications is not always easy to measure due to the extremely low concen-

trations of Co and sometimes there are problems with high background levels due to

sources of Co in the sampling system (e.g. Stellite in sampling valves). A good way to

monitor the long term trend of Co inflow to the reactor circuit is to regularly perform fuel

scraping campaigns to access the accumulation of Co and other corrosion products in the

fuel crud.

One additional way to reduce Co inflow is to introduce improved methods to avoid intru-

sion of Co due to grinding of valve seats during outages. Such methods are strongly em-

phasized, being an important part of the overall Co reduction program.

The BWR power uprates studied in more detail, OL1/2 (Section 4.1) and CNC (Section

4.2) have not included major replacement of Stellite in valves. The experience of such

replacements is rather sparse. The Ringhals 1 plant has conducted a considerable reduc-

tion of Stellite in the recirculation loops and in the LP turbines during performed mod-

ernization projects, and a significant gradual reduction of Co-60 radiation has been ex-

perienced. Some reduction of feedwater Co has also been seen in the Swedish forward

pumped heater drain plants Forsmark 3 and Oskarshamn 3 after the replacement of Stel-

lite in the turbine plants.

5.1.3.1.3 NWC – HWC – NMCA – OLNC

The method employed to establish reducing conditions in parts of the BWR primary cir-

cuit is to inject hydrogen into the feedwater. This technique was first demonstrated in

1979 in the Swedish Oskarshamn 2 plant, and was introduced in several plants during the

80’s. More than 50% of all BWRs are today on HWC in various degrees. The most recent

EPRI water chemistry guidelines from 2004 specifies that BWR components must be

exposed to reducing conditions (i.e. <-230 mV (SHE)) to mitigate stress corrosion crack-

ing initiation and growth. According to EPRI this goal can be achieved in two different

ways:

1. By injecting enough hydrogen to the feedwater to maintain the electrochemical

corrosion potential (ECP) of reactor internals below -230 mV (SHE). This

method is defined as moderate hydrogen water chemistry (HWC-M).

2. By treating the reactor internals with noble metals (Pt and Rh) and injecting small

amounts of hydrogen into the feedwater to maintain a hydrogen to oxygen molar

ratio of ≥3 (which assures an ECP value <-230 mV (SHE) for reactor internals).

This method is called noble metal chemical addition (NMCA).

7 There are examples of small regulation valves where erosion of valve seats means a considerable loss of Co

material.

111

Most European plants, especially Scandinavian and German BWRs with internal recircu-

lation pumps, maintain operation on NWC. In most of the European plants, the reason for

this difference between plants is due to more resistant materials from environmental

cracking point of view The HWC technique relies on suppressing the radiolytic produc-

tion of O2 and H2O2, the oxidants responsible for promoting high ECP values on BWR

components, from values of several hundred ppb to ppb or sub-ppb levels. Because of the

complexity of the interaction between radiolysis production, removal of radiolysis prod-

ucts by boiling and decomposition, the concentration of oxidizing species (O2 and H2O2)

is not uniform around the circuit. The net result is that different components reach the -

230 mV (SHE) specification at different hydrogen injection rates. In addition, the hydro-

gen demand to maintain a given location below -230 mV (SHE) can change considerably

between different plants and during the cycle.

During the 1990’s a new concept capable of coping with many of the observed problems

with HWC was developed. This concept was a technique to transform all system surfaces

to respond readily and quickly to HWC and the reducing conditions as a noble metal re-

dox electrode. The increased response to hydrogen only demands a stoichiometric excess

of hydrogen, i.e. at least twice a much dissolved hydrogen as oxygen on the molecular

basis dissolved in the water, as opposed to the sometimes several hundred times higher

hydrogen over-stoichiometry required in ordinary HWC. This new concept, called

NMCA, Noble Metal Chemical Addition, involves an addition of noble metal, NM, com-

pounds, i.e. directly to the primary circuit. The NMCA will lead to NM deposits by an

electro less method in situ, resulting in the requested behaviour, as shown in Figure 5.1.6.

Figure 5.1.6: Schematic behaviour of stainless steel and platinum in BWR water

as a function of feedwater hydrogen addition, Hettiarachchi, S, 2000 [5]

112

The number of plants having applied NMCA or NobleChem™ has grown drastically

during the last few years. Recently, 24 U.S. BWRs (70% of the U.S. BWR fleet), 1 Euro-

pean, and 3 Japanese BWRs have applied NMCA. Hence 28 of 88 BWRs worldwide

apply NMCA. Fifteen plants have operated for at least two cycles with NMCA, but only

one plant has finished more than three cycles, and that is Duane Arnold. A couple of

plants have applied NMCA but do not also perform hydrogen addition, for various rea-

sons. If the rate of introduction of NMCA is compared with other important water chem-

istry adjustments in U.S. BWRs in the last two decades, i.e. HWC and zinc injection, see

Figure 5.1.7, it is obvious that much less experience was been gained from NMCA com-

pared to the other regimes.

Figure 5.1.7: Water chemistry changes in US BWRs, Jones, 2004

The water chemistry will be rather drastically affected by the NMCA, as has been re-

corded to at least to some degree in all cases where NMCA has been employed. One ef-

fect is an initial increase in the MSLR levels. This increase has been found to be up to a

100% increase in the MSLR even at NWC, and occasionally equally high transients are

obtained once the initial increase at NWC has declined and HWC is started. The transient

can last from a few weeks to several months as described more in detail in a later section.

Another effect is the increase in conductivity following NMCA. This increase in conduc-

tivity can be rather significant. It was e.g. reported in Peach Bottom-3 that the maximum

soluble iron level was 44 ppb during the transient. If it is assumed that the iron was in the

form of (Fe2+

+ 2 OH-), a conductivity of 0.407 would be expected, which, at least in this

case, fully explains the conductivity transient. The largest conductivity transient reported

was of peaking at 1.23 µS/cm. This corresponds to 140 ppb of fully dissociated Fe(OH)2.

If, as seem to be the general case, the increase in conductivity is from soluble iron it will

probably not have any effect on cracking, as opposed to conductivity enhancements

caused by sulphate or chloride transients. Nevertheless, an elevated conductivity in the

range of 0.3 to 1.2 µS·cm-1

could conceal significant additional conductivity transients

113

caused by the more detrimental ions. It is recommended to operate the plant with both

zinc injection and HWC before NMCA in order to minimize the iron restructuring.

Another chemistry effect that seems coupled to the iron restructuring, is the release of

tramp uranium from system surfaces depositing on the fuel cladding surfaces. Other

chemistry effects appear to be more directly related to the establishment of a reducing

environment, as the enhanced iodine volatility and the MSLR increase, described above.

There are also a number of other water chemistry effects that are less straightforward to

understand. A summary of different water chemistry effects, and their grouping in dura-

tion, is presented in Table 5.1.1.

114

Table 5.1.1: Tentative grouping of post-NMCA chemistry effect duration.

Duration Effect

MSL radiation increase

Short-lived nuclides increase

Elevated conductivity Months

Elevated soluble iron

Elevated Co-60

Years Lowered sulphate

(and large hide-out return transients)

Tramp uranium vagabonding

Elevated offgas Long term

Lowered RW iodine

A summary of the radiation behaviour of seventeen plants that started hydrogen injection

immediately after NMCA is shown in Figure 5.1.8 where the BRAC standard dose rate

locations measured after the first and second cycles are plotted versus the cycle median

reactor water Co-60(s)/Zn(s) ratio. Seven of the plants were below the EPRI recom-

mended ratio of 0.74 Bq/ml/ppb for both cycles and their BRAC dose rates remained in

the less than 1 mSv/h range. Several of the plants experienced higher dose rates after

operating well above the 0.74 Bq/ml/ppb ratio value during the first post NMCA cycle

followed by a decreased ratio during the second cycle. These results clearly indicate that

those plants that operate their first post NMCA cycle at values below the 0.74 ratio will

have low dose rates as long as they stay at or below that value. For those plants that oper-

ate their first post NMCA cycle at values above the 0.74 ratio will “lock in” higher dose

rates and will probably have to conduct a piping decontamination and reapplication and

then operate at or below the 0.74 ratio to reach BRAC levels of 1 mSv/h or lower. Three

plants have utilized this strategy of conducting a piping decontamination followed by a

low temperature NMCA application.

115

µµ µµS

v/h

µµ µµS

v/h

Figure 5.1.8: Post NMCA BRAC dose rate measured at the end of the first and second

fuel cycles with highest dose rate plant omitted (Data available as of Jan. 26, 2006 - Post

NMCA Zinc Demand) [7]

Analysis of BRAC shutdown dose rates of plants that use moderate HWC chemistry and

zinc injection has revealed an apparent grouping based on the recirculation piping elec-

trochemical corrosion potential (Figure 5.1.9). All plants shown (with the exception of

two), inject zinc according to the recommended reactor water soluble Co-60 to soluble

zinc ratio of 0.74 Bq/ml/ppb. The data reveals two clear groupings between the plants.

Group 1 plants enjoy average BWR Radiation Assessment and Control (BRAC) dose

rates similar to those experienced by 1st post NMCA cycle plants that follow the recom-

mended cobalt-to-zinc ratio, while Group 2 plants have considerably higher dose rates.

Cowan and Hussey, 2006 [7] conclude, that the apparent difference between Group 1 and

Group 2 plants is the electrochemical corrosion potential (ECP) in the recirculation pip-

ing. Although in both groups the recirculation piping ECP is below -230 mV SHE (Stan-

dard Hydrogen Electrode) and has sufficient mitigation protection against intergranular

stress corrosion cracking (IGSCC), the recirculation piping from Group 1 plants has an

ECP in the range of the theoretical limit from hydrogen injection. Recall that NMCA

plants effectively have theoretical ECP at the surface in the presence of small amounts of

hydrogen. Therefore, the trends are consistent; if the recirculation piping surfaces main-

tain the theoretical hydrogen ECP, then the plant should observe similar radiological

benefits. This is further supported by the decreased BRAC shutdown dose rates for Grand

Gulf. The figure shows the transition from higher dose rates to lower dose rates after the

hydrogen injection rate was increased.

116

µµ µµS

v/h

µµ µµS

v/h

Figure 5.1.9: Grouping of HWC plants that inject zinc based on recirculation piping ECP [7]

Cowan and Hussey, 2006 [7] also discuss the influence of feedwater iron on shutdown

dose rates for plants that inject zinc. Iron would appear to have a direct impact on the

shutdown dose rates, but is actually secondary to its impact on zinc. Iron inhibits the per-

formance of zinc injection in three ways; (1) it acts as a competing metal for incorpora-

tion into the oxide film, (2) it may combine with zinc in the reactor coolant and prevent

zinc incorporation into the corrosion film, and (3) fuel performance concerns limit the

total metal oxide concentration in the reactor coolant, hence the more iron in the feedwa-

ter, the less zinc can be used. Iron reduction strategies were discussed in an earlier chap-

ter. The overall message is that reduction of feedwater iron is a very efficient way to re-

duce radiation fields in plants on HWC/NMCA together with DZO.

The overall conclusion is that low radiation fields can be maintained in HWC and NMCA

plant if the above recommendations are followed. Whilst these recommendations are not

directly affected by a power uprate they may be easier to realise if considered within a

plant modernisation project.

5.1.2.1.4 Zinc injection

The recommendation to inject zinc in especially plants on HWC or NMCA has previously

been addressed. A recent example of a successful application of Zn injection is the Oskar-

shamn 2 plant. Of special interest is the experience of the combined effect of low feedwa-

ter iron, HWC operation and zinc injection in the Oskarshamn 2 plant, reported by Lejon,

2006 [8]. The HWC operation in the Oskarshamn 2 plant resulted in a considerable in-

crease of radiation levels, with an increase of RHR dose rate up to about 7 mSv/h at the

2003 refuelling outage (Figure 5.1.11) The increase in the dose rate was in spite of very

low feedwater iron and rather low reactor water Co-60 level (Figure 5.1.11), and a main

reason for the increase was believed to be due to a significant contribution of soluble iron

from certain reactor systems reaching very low corrosion potential because of the HWC

117

operation. A decision was made to perform a system decontamination during the 2003

outage, and thereafter start DZO injection during the following cycle. The DZO injection

resulted in an increase of reactor water Co-58 and Co-60 (Figure 5.1.10), but the RHR

dose rate has remained on a low level (Figure 5.1.11).

Oskarshamn 2Co-58 and Co-60 in RW

0,0E+00

5,0E+03

1,0E+04

1,5E+04

2,0E+04

2,5E+04

3,0E+04

3,5E+04

4,0E+04

4,5E+04

5,0E+04

jan-01 maj-02 sep-03 feb-05 jun-06

[Bq/kg]

DZO

Co-58

Co-60

Figure 5.1.10: Oskarshamn 2 – Reactor water Co-58 and Co-60 before and after DZO

injection

O2 - BetongOla - Detektor 1 - Dosratsbidrag

0

1

2

3

4

5

6

7

8

9

10

jan-02 apr-02 jul-02 okt-02 jan-03 apr-03 jul-03 okt-03 jan-04 apr-04 jul-04 okt-04 jan-05 apr-05 jul-05 okt-05 jan-06 apr-06 jul-06

DR

[m

Sv/h

]

0

20

40

60

80

100

120

Eff

ekt

[%]

D1-N16

D1-O19

D1-Crud

Effekt

Figure 5.1.11: Oskarshamn 2 – On-line follow-up of contact dose rate on RHR pipe

with separation of different radiation sources (DZO injection started Oct. 2003)

The water chemistry conditions in Oskarshamn 2 during DZO injection can be summa-

rized as follows [8]:

• Feedwater Zn: about 0.2 ppb

Reactor water zinc: 4-5 ppb

- i.e. about 50% of injected Zn cleaned up in RWCU (2% RWCU capacity)

• Feedwater Fe: typically 0.02 ppb

118

• HWC operation:

- PLR piping: about -300 mV (SHE)

- RHR piping: about -500 mV (SHE)

• Fuel crud composition in fuel crud that has only seen DZO conditions:

- (Fe/(Ni+Zn)) ~ 2.2, see Table 5.1.2.

The outcome in Oskarshamn 2 coincides quite well with the proposed optimized corro-

sion product balance described previously. The overall corrosion product balance must be

addressed in any power uprate project.

Table 5.1.2: Oskarshamn 2: Fuel deposit sampling data one cycle after DZO introduction

[8]

5.42.2[Fe/(Ni+Zn)]

7.31.7[g/m2]Σ

<0.01<0.01[g/m2]Cu

0.0130.005[g/m2]Co

0.150.05[g/m2]Mn

0.140.14[g/m2]Cr

0.190.31[g/m2]Zn

0.970.22[g/m2]Ni

5.91.0[g/m2]Fe

DZO 1 of 5 DZO only

373377942[EFPH]

41.310.7[MWd/kgU]Burn-up

Fuel assembly

5.42.2[Fe/(Ni+Zn)]

7.31.7[g/m2]Σ

<0.01<0.01[g/m2]Cu

0.0130.005[g/m2]Co

0.150.05[g/m2]Mn

0.140.14[g/m2]Cr

0.190.31[g/m2]Zn

0.970.22[g/m2]Ni

5.91.0[g/m2]Fe

DZO 1 of 5 DZO only

373377942[EFPH]

41.310.7[MWd/kgU]Burn-up

Fuel assembly

5.1.2.2 Radiation fields during operation

The radiation fields during operation are dominated by the contribution from the short-

lived radionuclide N-16 (T½ = 7.13 s). Most of the produced N-16 is decaying in the reac-

tor vessel and recirculation lines, this activity is well shielded and does not contribute to

the occupational exposure. However, the carry-over of N-16 to the turbine plant is of

great importance, and particularly the resulting radiation levels outside the turbine plant

due to air-scatter radiation through the roof of the turbine building (s.c. N-16-skyshine).

The radiation source term of N-16 activity in the turbine plant is affected by a power

uprate in two different ways:

1. The power uprate in itself increases the production of N-16 in the reactor coolant

in proportion to the power increase.

2 The increased steam flow rate in association with the power uprate means a re-

duced transport time from the reactor core to the turbine plant. A typical transport

time is about six seconds, and e.g. a 30% power increase with a corresponding in-

crease of steam flow rate means a reduction in transport time of about 1.4 s. The

overall effect of such a power increase would be about a 50% increase of the N-16

inventory in the turbine plant.

119

On the other hand, such an increase of N-16 activity in the turbine plant is much less

dramatic than the increase experienced when HWC is introduced, when the N-16 carry-

over to the steam is typically increased by a factor of five. Therefore, the biggest concern

from the perspective of N-16 skyshine, is for plants combining HWC operation with a

significant power uprate. It has also to be noted, that different plants have different well

shielded turbine plants. In practice the difference in N16 skyshine levels outside a well

shielded, compared to less effectively shielded, plant under the same operational condi-

tions can be almost an order of magnitude. Almost all BWRs were originally designed for

NWC operation, i.e. HWC operation is not considered in the shielding design.

Nevertheless, the power uprates studied in both the OL1/2 (Section 4.1) and CNC (Sec-

tion 4.2) plants show only marginal impact on the occupational exposures during reactor

operation, and these exposures remain ..only a small fraction of the total exposures. The

OL1/2 plants are on NWC, while the CNC plant is on HWC operation.

5.1.2.3 Radiation fields during outages

As discussed in the previous sections, the radiation fields during outage conditions are

mainly controlled by the overall corrosion product balance. This corrosion product bal-

ance is affected by a power uprate, but correctly addressed it can be maintained and even

improved after the uprate. This is verified by the experience of the plants in the present

study Section 4.1 and Section 4.2.

The normal BWR case is low steam moisture content and low shutdown radiation levels

in the turbine plant. However, there is experience of situations with increased steam mois-

ture, resulting in significant radiation levels and occupational exposures in the turbine

plant (see e.g. the OL1/2 plants Section 4.1). Such situations of increased steam moisture

content are in many cases associated with plant power uprates. Critical reactor compo-

nents determining the steam moisture content are the steam separators above the core plus

the steam dryer in the top of the reactor pressure vessel. These components are sensitive

to the core and steam flow, and the increase of flow rate in association with power uprates

may result in a break-through of moisture. The lesson learnt is that the design of steam

separators and dryers must be carefully reviewed when planning an uprate, and that sig-

nificant redesign of these components may be necessary.

5.1.2.4 Fuel failure issues

During the past three years the number of US BWRs with cladding defects has doubled,

so that today about one-third of the US plants are operating with at least one leaking fuel

rod. Understanding and reversing this trend is a top industry priority because fuel clad-

ding failures lead to increased O&M costs, as well as to operational restrictions and out-

age duration increase.

Determining the root causes of the recent failures is complicated by the fact that fuel de-

signs and duties, cladding materials and primary water chemistries have all changed sig-

nificantly during the past decade. The root cause evaluations performed indicate that a

considerable fraction of recent failures is associated with water chemistry and crud depo-

sition, see Figure 5.1.12. However, it is necessary to remember that a large fraction of the

crud related failures seen in the figure is from one single plant with a large number of

failures. Research programs have been initiated to gain better understanding of the

mechanisms, and possible identify what water chemistry changes shall be introduced to

minimize the formation of deleterious fuel crud. It should be observed that the increase of

120

fuel crud related failures coincides with the extensive introduction of changed chemistry

conditions in the form of zinc injection and NMCA, see e.g. Keys, 2004 [9]. The consid-

erable effort in US to reduce feedwater iron is, to a large extent, driven by the aim of

reducing crud induced fuel failure problems.

A similar trend of increased fuel crud associated fuel failures are not seen in Scandina-

vian and German plants, further indicating that the main root cause of the US experience

is due to the significant change of water chemistry conditions introduced during last

years.

Figure 5.1.12: BWR fuel failure root cause trend in US BWRs, Turnage, 2004 [1]

Fuel failures, including dissolution of fuel material, result in increased release of fuel crud

particles to the coolant, which in turn results in the formation of radiation sources (hot

spots) in the plant. The responsible mechanism is knock-out reactions due to fissions in

the formed tramp uranium. A significant tramp uranium contamination results in long-

term effects, which take about 10 years to return to the background level. Plant experi-

ence has shown that a tramp uranium contamination of the order 50 – 100 g may increase

the annual exposure with typically 20%. There are also several other negative impacts of

fuel dissolution resulting in tramp uranium on the core:

− High background activity levels in reactor water and off-gases making it more

difficult to detect additional fuel failures.

− Increase of release to the environment.

− Increased waste production, especially of actinides.

− Increased radiation levels around certain turbine components due to accumulation

of noble gas daughters (e.g. Ba-140 and La-140).

The way to avoid this problem is to introduce a proper fuel failure management at the

plant. Methods should be introduced to efficiently monitor situation with fuel dissolution.

The earlier method using measurements of reactor water Np-239 has been shown not to

121

work in HWC and NMCA conditions. Therefore, many plants have introduced a method

where Sr-92 reactor water activity is used as a measure of the amount of tramp uranium.

This method is efficient both in oxidizing and reducing conditions. The employed method

must be combined with a determined level of maximum amount of accepted tramp ura-

nium. Most Scandinavian BWRs have introduced such a level, typically in the order of 10

g total of tramp uranium. In situations where there is a risk that this level will be ex-

ceeded the reactor would be shut-down, the leaking fuel identified and removed.

Both the OL1/2 (Section 4.1) and CNC (Section 4.2) plants have recently experienced

failures of the foreign debris type. Several plants take actions to reduce the risk of fuel

failures of the debris type by introducing fuel with debris filters and by installing cyclo-

tron filtration in the main feedwater lines. The crud induced failures recently seen in some

US plants are the results of bad feedwater chemistry (high levels of Fe, Zn and Cu), and

actions are underway to improve the situation, see e.g. Figure 5.1.3.

5.1.2.5 BWR conclusions

− BWR plants worldwide continue to show low annual occupational exposure,

typically 1 – 2 manSv per plant and year. The lowest exposures are experienced

in BWR designs with internal recirculation pumps. No significant difference be-

tween plants being power uprated or non uprated is seen.

− A key factor in achieving low radiation levels seems to be to maintain a well bal-

anced corrosion product inflow to the primary circuit, forming low amounts of

fuel crud of the spinel type. Efforts have to be spent to improve condensate

cleanup, especially in connection to power uprates, in order to achieve that. Zn

injection seems to be an effective method to reduce radiation fields in plants on

HWC or NMCA.

− NMCA plants that have already maintained the recommended reactor water zinc

level during the first NMCA cycle show low piping dose rates still after two cy-

cles, around 1 mSv/h. For those plants that operate their first post NMCA cycle at

lower Zn levels will “lock in” higher dose rates and will probably have to con-

duct a piping decontamination and NMCA reapplication to reach dose rates of

1 mSv/h or lower.

− HWC plants that are operating with the piping ECP levels close to the theoretical

for Pt (<-450 mV (SHE) show lower radiation fields than plants operating closer

to the -230 mV (SHE) limit. The “Low-ECP” HWC plants show radiation fields

at about 1 mSv/h, i.e. similar to the NMCA plants.

− European BWRs show similar radiation levels to US BWRs in spite of some dif-

ferences in reactor design and water chemistry conditions. One Scandinavian

BWR demonstrates that low radiation fields can be achieved with the combina-

tion HWC, Zn injection and ultra-low feedwater Fe (<0.1 ppb).

− In many cases increased steam moisture content is associated with plant power

uprates. Critical reactor components determining the steam moisture content are

the steam separators above the core plus the steam dryer in the top of the reactor

pressure vessel. The lesson learnt is that the design of steam separators and dryer

must be carefully reviewed when planning an uprate, and that significant redesign

of these components may be necessary.

122

− Several plants take actions to reduce the risk of fuel failures of the debris type by

introducing fuel with debris filters and by installing cyclotron filtration in the

main feedwater lines. The crud induced failures recently seen in some US plants

are results of bad feedwater chemistry (high levels of Fe, Zn and Cu), and actions

are underway to improve the situation.

5.1.3 PWR uprates

5.1.3.1 Water chemistry issues

The primary side reactor water chemistry in a PWR has a large impact on the activity

buildup, and hence on the radiation fields and occupational exposures. The chemistry is to

a large extent determined by the power regulation with boron, and a power uprate will

certainly affect this boron control. The influence of a power uprate on the water chemistry

conditions, and the resulting impact on radiation levels, are discussed in this section.

5.1.3.1.1 Boron

The reactor power, and the corresponding variation of boron in the reactor water during

the period 1995 – 2005 in the Swedish PWR Ringhals 3 (R3), is shown in Figure 5.1.13.

A boron concentration of at least 2000 ppm is maintained during outage conditions to

assure under-criticality. The boron content is lower at power operation, but in the begin-

ning of a fuel cycle (BOC) considerable boron content is needed to compensate for the

over-reactivity of the newly loaded core. The boron content at BOC varies somewhat due

to how long the fuel cycle and how long the coast-down operation are planned for. The

boron cycle is almost linearly decreased during the cycle down to 3-4 ppm, when coast-

down operation is started.

A typical recent fuel cycle in R3 is shown in Figure 5.1.14.

123

R3 - Boron in RC

0

250

500

750

1000

1250

1500

1995 2000 2005

B [

pp

m]

0

20

40

60

80

100

120

Po

wer

[%]

B

Power

Figure 5.1.13: R3 – Boron reactor water concentration and reactor power 1995 - 2005

R3 - Boron in RC

0

250

500

750

1000

1250

1500

maj-04 jun-04 jul-04 aug-04 sep-04 okt-04 nov-04 dec-04 jan-05 feb-05 mar-05 apr-05 maj-05

B [

pp

m]

0

20

40

60

80

100

120

Po

wer

[%]

B

Power

Figure 5.1.14: R3 – Boron reactor water concentration and reactor power during the fuel

cycle 2004-05

An increase of power level means that the BOC boron content must be increased. This is

exemplified in Figure 5.1.15, where an increase from 2873 to 3160 MWth is assumed in

the R3 plant. The BOC reactor water boron is increased from 1250 to 1460 ppm, and the

average boron conten from 634 to 741 ppm. Note that longer fuel cycles mean a further

124

increase of average boron content, as illustrated by data from a PWR operating with 18

months cycles (Figure 5.1.16).

R3 - Boron in RC

0

250

500

750

1000

1250

1500

maj-04 jun-04 jul-04 aug-04 sep-04 okt-04 nov-04 dec-04 jan-05 feb-05 mar-05 apr-05 maj-05

B [

pp

m]

0

20

40

60

80

100

120

Po

wer

[%]

B

Power

Figure 5.1.15: R3 – Typical variation of reactor water boron at 2873 to 3160 MWth

Figure 5.1.16: Variation of reactor water boron in a plant operating with 18 months cy-

cles

125

The main impact of the boron content is on pH control and on the risk of being affected

by Axial Offset Anomaly (AOA). Both these issues are discussed further in later sections.

The content of boron also affects the production of tritium (H-3, T½ = 12.33 y) in the

coolant. Whilst the production occurs through several reactions it is dominated by the

following:

• B-10 (n,2α) H-3

The combination of increased neutron flux and increased average boron content in the

reactor water means that the production of H-3 will increase alsont four time with the

reactor power increase. The increased H-3 production will only marginally affect the

radiological in-plant conditions, but has a significant impact on the activity release from

the plant.

The corresponding quadratic increase with reactor power will be experienced for the ra-

dionuclide carbon-11 (C-11, T½ = 20.385 m):

• B-11 (n,p) C-11

C-11 has a rather small effect on radiation levels during operation and activity release

from the plant.

So called EBA (Enriched Boron Acid), where the fraction of the neutron absorbing nu-

clide B-10 has been increased, is used in some plants. EBA with typically 30% B-10

compared to 19.7% in natural boron is used, which means that the reactor water boron

content can be reduced with about 30%. The use of EBA has advantages in pH control

and to avoid AOA, see below. Furthermore, the production of C-11 is reduced. However,

the production of H-3 in the coolant is not affected due to a maintained content of B-10.

5.1.3.1.2. pH control

Adding boron acid to the reactor coolant means acid conditions, which is counteracted by

adding 7-litiumhydroxide (7LiOH) to the coolant. The selection of

7Li as counter-ion to

the hydroxide is made from materials and reactor physics considerations8. Early operation

in US PWRs was made with constant concentration of 7Li in the coolant, which resulted

in a gradual increase of alkaline conditions during the fuel cycle. This operation resulted

in high radiation fields around steam generators (SGs) and primary piping, the main rea-

son associated with corrosion and solubility of corrosion products. The corrosion rate of

SG material was high at the BOC conditions with less alkaline conditions, and the corro-

sion products were predominantly deposited on the hot fuel surfaces. The more alkaline

conditions at the end of cycle (EOC) conditions resulted in a dissolution and redistribu-

tion of the neutron activated fuel crud to the colder SG tubes and primary piping.

The solution to that problem was the introduction of so called “coordinated” chemistry,

where the content of 7Li in the reactor water is varied in order to maintain a constant high

temperature pH (pH300) during the cycle. However, the question was what pH300 to se-

lect? Initially the solubility of magnetite was selected as a base for the decision, resulting

in a selection of pH300 = 6.9. Later it was realized that the PWR chemistry was more af-

fected by the solubility of nickel oxides, especially in the case of SG tubes of Inconel 600

8 7Li is also produced in neutron absorptions in 10B. 7Li has a low neutron cross section as compared to 6Li,

which means a low production of H-3. Added 7Li is specified with a low level of 6Li (<0.04%).

126

or 690. Therefore, an effort has been made to introduce operation at increased pH300, up

to 7.4 during most of the cycle (”Elevated” pH chemistry), where the nickel oxides seem

to have minimum solubility. The operation experience with such elevated pH chemistry

has generally been favourable, with reduced radiation levels during outages around SGs

and primary piping.

The relation between content of boron and Li for some high temperature pH is shown in

Figure 5.1.17. The figure shows that a considerable Li concentration is needed to main-

tain a pH300=7.4 at BOC. A sensitive question in the PWR society is the maximum level

of Li in the coolant that is acceptable for the fuel cladding. A maximum level of 2.2 ppm

has been adopted for a long time [11]. This limit, however, means that the high tempera-

ture pH is low during a large part of the beginning of the cycle, and the optimum

pH300=7.4 is only obtained at the end of the cycle. The fuel vendors have through materi-

als development allowed higher maximum Li concentrations, and the limit used during

several years at the Ringhals PWRs is 3.5 ppm. Even at Limax=3.5 ppm a large fraction of

the fuel cycle is not operated at a pH300<7.4, especially in the case of uprated power with

increased boron concentration. Presently projects are underway to increase the maximum

Li content in US and European BWRs, and the Ringhals PWRs have recently achieved

approval from the fuel vendor to increase the maximum Li limit to 6 ppm. It is used in R2

and R3 PWRs to be able to start at BOC at a pH300=7.2, and that Li content is maintained

up to pH300=7.4, thereafter pH300=7.4 is maintained by adjusting the Li content.

The conclusion is, that with an allowed increased maximum Li level, the pH control may

be maintained or even improved after a power uprate. The threat against such an opera-

tion strategy is the risk that future fuel experience would reveal that the maximum Li

content should be decreased, however such a development is not likely and not supported

by recent fuel inspections. One possible solution to handling such a situation is to intro-

duce EBA with reduced boron content and thereby lead to a reduced need of Li in the

coolant. Another possibility is to introduce Zn injection in the PWR primary circuit. Zn

injection in PWRs has been introduced in some US and German PWRs with favourable

impact on the radiation buildup.

127

0,0

1,0

2,0

3,0

4,0

5,0

6,0

7,0

8,0

9,0

10,0

0

100

200

300

400

500

600

700

800

900

1000

1100

1200

1300

1400

1500

1600

1700

1800

1900

2000

Li

(ppm)

Boron (ppm)

7,40

7,10

7,00

6,90

7,20

7,30

Figure 5.1.17: Relation between boron and Li for pH300 = 6.9, 7.0, 7.1, 7.2, 7.3 and 7.4 [10]

5.1.3.1.3 Corrosion products

The corrosion products in PWR primary circuit is to a high degree determined by the

selection of material in the SG tubes, in combination with the pH control applied. The

selection of Inconel 600 or 690 as tube material will mean the chemistry is very much

determined by nickel and its oxides. The corrosion product chemistry will mainly be de-

termined by these three factors working together:

1. The materials selection of the SG tubes. The present widespread use of Inconel

690 shows excellent corrosion performance, especially on the secondary side.

German PWRs instead use Inconel 800, which seems to have some advantages on

the primary side due to its higher content of Fe compared to the Inconel 690 (and

Inconel 600).

2. The pH control adopted. A pH300=7.4 seems to be close to optimal in the case of

Inconel 690 and 600 due to low solubility for nickel oxides.

3. The amount of dissolved hydrogen (DH). A typical value of about 35 ml

(STP)/kg used in most PWRs seems to be quite enough to ensure reducing condi-

tions in the whole primary circuit. However, the optimum DH concentration is

subject to discussions, and both higher and lower values than present are pro-

posed. The present guidelines may be changed in the future.

Measured reactor water Ni concentrations during normal operation and shutdown tran-

sients in the Ringhals PWRs are shown in Figure 5.1.18. The shutdown transients have

been followed during several years, while the normal operation concentrations have only

been followed during recent years through the introduction of integrated sampling tech-

niques. A couple of interesting observations can be made:

128

• The normal operation levels are low, about 0.1 ppb. Some US PWRs that have ex-

perienced fuel AOA problems have experienced considerable higher levels, up to 10

ppb [13].

• The shutdown transients show a considerable reduction in the R2 and R3 plants after

the increase of high temperature pH to 7.4 that was introduced around 2000. This is

further shown in Figure 5.1.19, where the accumulated annual amount of Ni in tran-

sients in the R3 plant is shown. The increased pH introduced after the 1999 outage

has resulted in a dramatic decrease in Ni transient, and a similar development is seen

in the R2 plant with the same pH control, but not in the R4 plant that has maintained

the old pH control with a somewhat lower pH.

NiRC - Measured R2/3/4 data

1.E-03

1.E-02

1.E-01

1.E+00

1.E+01

1.E+02

1.E+03

1.E+04

1.E+05

1990 1995 2000

[pp

b]

R2_g

R2_i

R3_g

R3_i

R4_g

R4_i

Figuer 5.1.18: R2/3/4 – Measured reactor water Ni during normal operation and shut-

down transients (g – grab samples, i – integrated sampling)

Normal

operation

ca 0.1 ppb

Decreasing

transients

129

Ringhals 3 Nickel accumulated removal during clean-up, cooldown [kg]

0.00

0.50

1.00

1.50

2.00

2.50

3.00

3.50

1996 1997 1998 1999 2000 2001 2002 2003 2004 2005

[kg

]

Nickel acc removal at cooldown kg

R3 changed maximum

litiumcontent from 2200 ppb to

3000 ppb after outage 1999

R3 changed pH[300] from 7.24

to 7.4, after outage 2000. R3

also changed max. litium

content to 3500 ppb 2000

New Steamgenerators at R3

After outage 1995

Figure 5.1.19: R3 – Evaluation of release of Ni during shutdown transients 1996 – 2005

[12]

5.1.3.2 Radiation fields during operation

The radiation levels during operation are very much determined by prompt radiation from

the core fission process, and the production of short-lived radionuclides such as N-16.

The increase is approximately proportional to power increase. Experience from operating

plants shows a rather small contribution to occupational exposure from these sources.

5.1.3.3 Radiation fields during outages

Radiation levels during outage conditions in PWRs are dominated by the nuclides Co-60

and Co-58. The radiation levels in plants that have adopted improved water chemistry

control during recent years, e.g. R2 and R3, normally show reducing radiation levels, see

Figure 5.1.20. On the other hand, plants that have maintained previous water chemistry

with somewhat lower pH, e.g. the R4 plant, show a less favourable development.

130

Ringhals PWR: 313 loops - Average shutdown radiation levels

0

0.5

1

1.5

2

2.5

1980 1985 1990 1995 2000 2005

DR

[m

Sv

/h]

R2

R3

R4

Ringhals PWR: SG Water Chambers - Average shutdown radiation levels

0

20

40

60

80

100

120

140

160

180

1980 1985 1990 1995 2000 2005

DR

[m

Sv

/h]

R2

R3

R4

Figure 5.1.20: Ringhals PWR – Dose rate trends on primary piping and in SG water

chambers

5.1.3.4 Fuel failure issues

5.1.3.4.1 Axial Anomaly Offset (AOA)

Axial Offset (AO) is defined as the relative power difference between the upper and

lower part of the reactor core:

Eq. 4 100t b

t b

P PAO

P P

−= ⋅

+

where:

Pt – Power in the upper part of the core

Pb – Power in the lower part of the core

131

The AO is calculated and measured during each cycle, and must be kept within specified

limits. If there is a difference between measured and calculated AO of at least 3%, an

Axial Offset Anomaly (AOA) exists. An example of AOA in a Westinghouse PWR is

shown in Figure 5.1.21. The beginning of the cycle shows a reasonable agreement be-

tween measured and calculated AO <3%. In the middle of the cycle the measured AO is

low compared to the calculated value, changing to a high measured AO compared to cal-

culation at the end of the cycle.

Figure 5.1.21: Example of AOA in a Westinghouse PWR [11]

All details for the AOA phenomenon is not well know, but the mechanism can qualita-

tively be explained in the following way (see also Figure 5.1.22):

− Ni-rich fuel crud is accumulating in the beginning of the cycle in high-power fuel

bundles. This is promoted by the coexistence of several factors:

- High contents of reactor water Ni, e.g. due to initial corrosion of a new SG

(with Inconel 690 or 600, AOA is never observed in German BWRs with Inconel

800 SG tubes) .

- Less good pH control due to high initial boron and too low Li. This problem is

especially pronounced in plants operating on long cycles, 18 or 24 months.

- High-power fuel assemblies with considerable sub-cooled boiling in the top.

The degree of boiling in the top of the core is normally substantially increased in

a reactor which has undergone power uprate.

− Boron is enriched in the formed fuel crud layer through boiling. The enrichment

of boron results in a not-predicted local power decrease.

− Dissolution of the fuel crud occurs at the end of the cycle, e.g. at power reduc-

tions. The dissolution of the fuel crud results in local increase of the power level

due to lower burnup than predicted due to the earlier boron accumulation.

132

− A large fraction of the accumulated fuel crud will be dissolved during the shut-

down transient. The dissolution of the fuel crud contributes to the activity buildup

and the shutdown radiation fields.

BOC MOC At outage

Temp

Boiling

Steam void

Crud deposit • High nickel content

10B co-deposition

Shut-down operation removes crud

MOC Lower power

Crud release

10B dissolution

Figure 5.1.22: Proposed mechanism for AOA

Significant problems with AOA have so far, only been experienced in US PWRs (long

cycles, maximum Li 2.2 ppm, Inconel 600 (690?), and high-duty cores). It is very obvi-

ous, that power uprates significantly increase the risk of AOA to be developed, something

that has to be addressed in an uprate project. Twelve months cycles and improved pH

control with pH300=7.4 are factors that reduce the risk. It may be a good decision to avoid

a power uprate immediately after a SG replacement when the release of Ni is expected to

be high due to the initial corrosion.

5.1.3.4.2 Other factors

A number of different types of fuel failures in PWRs are discussed in [14]. Some exam-

ples:

− Wear between spacers and cladding. One special case is so called “baffle jetting”

due to leakage through the baffle plates. These problems are avoided through im-

proved design.

− So called “debris fretting”. This failure mechanism is avoided by a high cleanli-

ness during outage, and by use of fuel with debris filters.

− Crud-related fuel failures have occurred in some plants. Is avoided by a good pH

control.

− Bending of guide tubes. This problem seems to be solved by improved fuel de-

sign and improved material selection.

133

− Pellet cladding interaction (PCI). More frequent in BWRs than in PWRs. Are

normally avoided in PWRs by restrictions on the rate of power changes.

− Pellet cladding mechanical interaction (PCMI). This failure mechanism is cou-

pled to the burnup of the fuel, and not primarily to the power level.

The performed review indicates, that the different failures have a very little connection to

the power level, and other factors, e.g. design and fuel material, are of larger importance.

5.1.3.5 PWR conclusions

− A power uprate in a PWR means increased content of boron in the coolant, espe-

cially in the beginning of cycle. One consequence is a quadratic increase of H-3

production. This effect is further enhanced if the plant is operated with long fuel

cycles (18 or 24 months).

− Operation with increased Li, with a high temperature pH around 7.4, has shown

to have a favourable impact on radiation levels. Earlier concerns of high-Li op-

eration with respect to fuel material have been reduced due to improved fuel ma-

terials. Improved pH control in most PWRs is the main reason for the present

rather low radiation fields and exposures in PWRs. Power uprates with increased

boron in the reactor water call for further increased reactor water Li in order to

maintain the favourable conditions. One alternative, however costly, is to intro-

duce EBA enriched in B-10 in order to reduce the boron content in the coolant.

− An additional way to achieve exposure reduction is to apply Zinc injection with

depleted zinc (DZO) also in PWRs. The effect of zinc on radiation fields seems to

be especially large in connection to SG replacements.

− One great concern for PWRs is the risk of having fuel problems due to so called

Axial Offset Anomaly (AOA). The development of AOA is promoted by the

combination of (i) high levels of Ni, e.g. after a SG replacement, (ii) poor pH

control with too low Li levels in BOC, and (iii) high power fuel assemblies with a

substantial sub-cooled boiling rate. The combination of power uprate and a SG

replacement means an increased risk, and it would be wise to avoid a power

uprate the first cycle after a SG replacement. Factors that reduce the risk is main-

taining a good pH control, and operating at shorter (twelve) months cycles.

− Other fuel failure mechanisms in PWRs seem to be mostly associated with fuel

design and materials factors, and only marginally with the power level.

134

5.2 Organisational factors

5.2.1 Reducing the exposures

Every NPP has stringent administrative processes and rules for performing the work that

has radiological impact on plant staff. Typically, these processes and rules apply to all

NPP staff. However, the rules are not always strictly followed, which might, in particu-

lar, be a problem with external (e.g. non-permanent) staff and contractors. Because of

that, the "administrative barriers" are not always effective as they are meant to be.

Moreover, the economic pressures resulted in refuelling and maintenance operations be-

ing streamlined and concentrated as much as possible, which may sometimes have a

negative impact on the occupational exposure.

The occupational exposures are best managed through appropriate job preparation, plan-

ning and implementation, and post-performance review. This is to ensure that exposures

are kept "as low as reasonably achievable" (ALARA). The use of operating experience

and exchange of best practices is an important prerequisite for applying the principle of

optimisation to occupational radiation protection.

NPPs worldwide have made significant progress over the past 10 years in reducing the

radiation dose to workers. The most important organizational elements in decreasing

exposures are:

5.2.1.1 Leadership and Supervisory Monitoring

− Recognition of the importance of radiation safety to all nuclear power plant workers

is an important issue for the nuclear industry and has to be communicated clearly.

Leaders need to emphasize the importance of the individual having a high awarness

of their personal responsibility in regard to radiation protection. This should include

the need to pay specific attention to complying with rules and following administra-

tive procedures and processes.

− The radiation protection management team has to be fully integrated into:

� Work planning activities

� Outage planning and scheduling activities

� Plant modification reviews

� Plant strategic decision-making processes.

− For those jobs where large doses could be received in a short period of time, di-

rect involvement by both radiation protection staff and work supervisors is re-

quired in:

� ALARA reviews

� Planning

135

� Preparation

� Performance of the job

Prejob briefings, review of prior operating experience, and adequate radiation exposure

reduction reviews are important measures required by ALARA planning. Radiation pro-

tection staff and work supervisors should ensure that necessary surveys and other radia-

tion protection measures required by ALARA planning are implemented at the work site.

5.2.1.2 High quality procedures

The activities with the potential for high radiation exposure have to be controlled by writ-

ten procedures and/or specific radiation work permits.

− Procedures or radiation work permits should contain specific radiation protection

instructions to prevent unplanned exposures.

− Procedures or radiation work permits should contain specific radiation protection

instructions such as permissible cumulative doses allowed for a task and, the requir-

ments relating to radiation monitoring (continuous or at what frequency, physically

or through remote monitoring methods).

− Procedures or radiation work permits should contain specific instructions concerning

the frequency and type of radiation surveys and at what dose rates or accumulated

dose to an individual worker the area has to be evacuated.

− Electronic dosimetry, with alarms for dose rate and cumulative dose appropriately

set for the task, should be required by procedures or radiation work permits and

when specified should be used.by personnel

− Entry into very high radiation areas should have the written approval from the radia-

tion protection manager.

5.2.1.3 Training

Important elements in reducing occupational exposures:

− Training of each ALARA activity

− Retraining of plant staff

− Training of contract employees in radiological protection expectations and practices

at NPP.

The training and retraining programs should stress the potential for abnormally high or

rapidly changing radiological conditions, including the actions required when these con-

ditions occur. Training should also emphasize the importance of a high level of aware-

ness and sense of individual responsibility with regard to personnel radiation protection.

136

5.2.1.4 Operation Experience

It was found that identifying and trending minor radiological protection violations and

problems that occur in an NPP helps to identify weaknesses, so that the defences and

barriers can be strengthened before leading to a more significant events.

5.2.1.5 Assessment of Radiation Conditions

− Periodic reviews and assessment of radiation protection deficiencies, including those

involving human performance errors assures that radiological controls are adequately

addressed. These reviews should ensure that work control systems, including radia-

tion work permits, procedures, and work orders, adequately address radiological con-

trols, including the potential for changing radiological conditions.

− Additionally, periodic self-assessments of radiation protection department perform-

ance should be performed to identify and correct radiation protection program and

process weaknesses.

− Review of plant areas should be performed to ensure that all areas with existing or

potential high radiation exposure rates are identified and properly posted and con-

trolled.

One important prerequisite for applying the principle of optimisation to occupational

radiation protection is the appropriate and timely exchange of data, techniques and ex-

perience on doses and dose reduction methods.

5.2.2 Implementation of power uprate

Many of the undesirable outcomes can be avoided if the uprate project organisation is

well resourced, staffed with personnel with extensive plant experience and focused on

identifying and resolving potential operational and other impacts. The elements, which

point out good organizational aspects, are:

5.2.2.1 Power Uprate Project Team Organisation

− The uprate project team includes a full-time project manager, and the project man-

ager and team members have no additional duties.

− Involvement of the operations, radiation safety and training departments early in the

project.

− Sufficient lead-time is assured to allow plant personnel adequate data analysis, iden-

tification of procedure changes, and reviews during all stages of the project.

− Sufficient time is assured for training plant personnel and for incorporating changes

in the control room simulator.

5.2.2.2 Feasibility Study Phase

− Feasibility studies considered existing equipment problems or limitations. Feasibility

studies that used original design parameters often underestimated the scope of work

required.

137

− Some project schedules can be challenged when detailed analyses identified major

additional equipment modifications.

− Following the feasibility study contributes to more effective reviews of power uprate

related plant changes, modifications, training, and procedure revisions

5.2.2.3 Detailed Analysis/Design Phase

− Plant engineering, radiation safety experts and operations personnel perform detailed

reviews of power uprate project analyses. Interdisciplinary occupation is very impor-

tant

− Ensuring that present plant conditions and equipment performance problems are

addressed during system and components design reviews.

− Insight into the critical topical areas - fuel

− Perform Loading Pattern Risk Assessment to address possibility of:

� Crud Induced Power Shift

� Steady state Hot-Leg Streaming

� Axial Xenon Stability

� Fuel Performance – Fuel Rod Design Criteria

� Axial Offset Anomalies

5.2.2.4 Implementation Phase

− For extended power uprates it is recommended to use a two-stage implementation

strategy to implement the changes; for example, new fuel design and power uprate

(good solution to avoid Axial Offset Anomalies after SG Replacement and power in-

crease)

− In the power ascension tests is included an approach in which power is increased in

increments, with hold points at predetermined intervals. These hold points can last

several days to gain experience or resolve technical issues at the new power level

− During the testing, actual plant parameters are compared to expected values

− A contingency plan is available to identify actions needed to address unexpected

plant situations.

5.2.2.5 Ongoing Post-Power Uprate Operation

− Staff are aware that operating margins are reduced and various plant systems and

components are changed

− Contingency plans are available in providing guidance to the plant staff when ex-

pected or unexpected conditions are encountered (vibration, flow accelerated corro-

sion, fuel condition)

138

− Following power uprates, some plants operate with equipment that was previously

used as spares. Removal of these components from service for maintenance follow-

ing power up-rates will require a power reduction from rated power. This has to be

correctly addressed because it has prompted changes to on-line maintenance strate-

gies.

− Power uprate projects sometimes require operating strategy changes. The traditional

operating philosophy of operating at 100 percent reactor thermal power may not be

applicable following many power uprates because a plant system or component may

be the power limiting factor.

− Ongoing equipment problems may occur due to flow-accelerated corrosion and vi-

bration. Because of greater feed-water and steam flows associated with power up-

rates, there is an increased potential for flow-accelerated corrosion that could lead to

failure. Additionally, increased vibration of components in systems experiencing in-

creased flow rates has caused fatigue-induced failures. These conditions may not be

readily identified during the analysis phase of the uprate project. Thus, the scope and

frequency of flow-accelerated corrosion and vibration monitoring programmes may

need to be reconsidered following power uprates.

Economic evaluations for adding electrical generation under present industry conditions

are likely to show justification for the continued power uprating of existing nuclear facili-

ties.

Industry experience has shown that power uprates can be implemented safely and suc-

cessfully. However, industry experience illustrates the need for additional focus on the

importance of using a thorough, deliberate approach when planning and executing power

uprates to avoid undesirable consequences.

5.2.3 Lesson learned

Leadership, composition and organisation of the power uprate especially large demanding

tasks are critical for successful implementation of power uprate and received doses. Les-

sons learned comprise different aspects:

− Human factor and work planning

Work in difficult conditions, like SG replacement, should be carefully planned and

supported with sufficient number of workers with the necessary skills. Otherwise,

underestimation could lead to: unnecessary injuries of the overloaded workers, job

extension with influence on the outage critical path and poor quality of the work per-

formed.

− Scheduling, documentation

Final documentation necessary for all modifications which are going to be performed

have to be presented at the site early enough, otherwise it could result in insufficient

139

time for the procurement of necessary parts and components and prolong the time of

outage.

− Project

Each large demanding task has to start sufficiently early in order to prevent the ac-

ceptance of inappropriate requirements due to time pressure. This can lead to accep-

tation of some deviations, which are normally not been accepted. More emphasis

should be put on early planning and involvement of interdisciplinary teams.

− Subcontractors

Most of the large activities require the involvement of subcontractors. They have to

be included in the project early enough for quality work preparation. Non-quality

performed work invariably led to additional work hours, which can result in in-

creased occupational exposures for both plant and outside personnel.

It has to be noted that it is not recommended to give responsibility for important

tasks to outside personnel unless there is strong supervision by plant personnel. Out-

side personnel are not necessarily as well-trained or acquainted with plant design de-

tails as plant personnel.

140

6. Conclusions

This report presents the result of the three tasks of the Inquiry into the radiological conse-

quences of power uprates at light-water reactors worldwide. The review has resulted in

the following conclusions:

Compilation of power uprates

Worldwide collection of information on the uprates for PWR and BWR reactors that were

implemented or are planned to be implement are summarised in the database. Through a

process of data collection and its review the following initial conclusions were obtained:

− Throughout the world, occupational doses at NPPs have steadily decreased over the

past decade, mainly through better application of ALARA principles, better use of

shielding material, but also increased attention to occupational dose issues.

− Occupational exposure in the BWR plants are typically about 50% higher than in

PWR plants, due to differences in the design

− No direct relationship between the uprates and the occupational doses could be estab-

lished. The occupational doses on some plants seem to be higher after the uprate,

while on others seem to be lower.

− There is no obvious correlation of the power uprate and fuel failures. However, per-

formance of fuel for PWRs and BWRs went in opposing directions, improving for

PWRs and deteriorating for BWRs.

− Through the data collection process events were identified that have occurred as a

result of inadequate design or implementation of uprates. These events involved

equipment issues, unanticipated responses to conditions, or challenges for operating

staff , for example:

� Loose parts as a result of a flow-induced, high-cycle fatigue failure on a

steam dryer cover plate (BWR plants)

� Operational transients and equipment damage due to lack of training of plant

staff on changes to PCS operating characteristics

� Unanticipated challenges and degraded performance from reductions in mar-

gins

� Operation beyond licensed power levels for extended periods due to errors in

thermal power calculations following uprates

− None of the above events had direct consequences on doses to the personnel or re-

leases. However, some of them might have had an indirect influence on occupational

exposure or releases (replace or repair of damage equipment).

141

Analysis of selected BWR plants:

Olkiluoto 1 and 2

The plants have been uprated twice since commissioning. The thermal power of each


The 1998 uprate was part of an extensive modernization program implemented in 1994–

2006.

- Good planning of modernization program has reasonable impact on outage lengths

(maximum annual outage length 22 days compared to typically 7-14 days).

- Investment in the cleanup capacity (still maintained after the power uprate) results in

favourable water chemistry conditions that can be maintained, or even improved, after

the power uprate.

- Reduction or replacement of materials (Stellite) results in Co source reduction.

- Exposures during operation are maintained on a constant and rather low level after the

uprate. One important factor is that the plant is maintained on chemistry without hy-

drogen injection.

- Radiation levels during outage on reactor systems are maintained on a rather low and

constant level after the power uprate.

- The installation of new steam separators can increase the radiation levels around main

steam lines and other turbine components due to a considerable increase in steam

moisture content.. This problem can be overcome with a recent design and installation

of new steam dryers in the reactor pressure vessel to reduce steam moisture.

- The exposure per outage day is maintained on a fairly constant level even though the

considerable man-hour efforts during some outages for the power uprate and plant

modernization program have resulted in increase of occupational exposures.

- The average annual exposures in the Olkiluoto plants is kept on a rather low level

compared to international BWR data in spite of the large efforts for power uprate and

plant modernization.

Cofrentes

The present power level corresponds to 111.8% of the initial thermal power level, which

means on average 5.19 MWt per fuel assembly. The main power increase was introduced

in 2002, when the power level was increased from 104.2% to 110%.

- Extension of the fuel cycle often goes in parallel with the power uprates. Due to the

margin in the core fuel assembly, design is changed from the original 8x8 array to 9x9

and finally to 10x10 necessary for the more demanding recent operation conditions.

- Modifications, which mainly affect the turbine plant, result in low exposure due to

low contamination level of the turbine plant.

- Increase in the reactor pressure and temperature due to uprate only moderately affects

the steam velocity in the reactor and the main steam lines.

142

- The reactor water chemistry is significantly influenced by the design and materials

selection of the turbine plant. Low-alloy steel pipes, instead of carbon steel, consid-

erably reduce contribution to the feedwater iron.

- Radiation fields are well controlled by the introduction of zinc injection

- Long term introduction of hydrogen water chemistry (HWC) results in reduction of

corrosion materials.

- The annual exposures during operation, affected by the production and distribution of

the short-lived nuclide N-16, are maintained on a rather stable level and do not seem

to be significantly affected by the change to HWC or the power uprates. Radiation

levels during operation indicate an average effect due to the power uprates in the order

+15% - +30%. This is probably due to the introduction of HWC with increased carry-

over of N-16. Overall it can be concluded that the N-16 radiation source term is not

the dominant contributor to the occupational exposure during operation.

- The recirculation loops and the RWCU piping significantly influence radiation levels

during outage conditions are of importance to occupational exposures

- An increase of the radiation fields around recirculation loops is experienced due to the

specific water chemistry situation (gradually decreasing reactor water copper and

HWC operation resulting in restructuring of the oxide layers inside recirculation

loops). Several measures are introduced to mitigate the increase, and the recirculation

loop radiation fields seem at present to be low and well controlled.

- The annual occupational exposures at CNC display a slightly increasing trend during

the last 10 years. This trend is explained by the combined effect of increased radiation

fields and the considerable modifications and maintenance that have taken place dur-

ing recent outages. A future decreasing trend is expected due to the above-mentioned

improved control of radiation fields around the recirculation loops.

- The CNC power uprates have had a negligible impact on occupational exposures, or at

least are shadowed by more important factors such as water chemistry.

- Analysis of selected PWR plants:

Asco

The plant has been uprated twice since the commissioning. The thermal power of the


The SG Replacement was performed in 1995 but all safety analyses necessary for power

uprate were performed in 2000. The present study has focused on the first uprate, result-

ing in a thermal power level of 8% compared to the initial power level. The latter uprate

was an uprate of 1.5 %, achieved by using more precise techniques for measuring feed-

water flow.

- SG Replacement significantly affected outage length and doses. The activity required

95 outage days. Typical outage lengths are between 30 and 40 days.

- Standard PWR water chemistry is maintained, no zinc is injected.

- The exposures during operation are maintained on a constant and rather low level after

the uprate.

143

- Dose rates, during outage, on reactor systems are maintained on a rather low and con-

stant level after SG Replacement.

- The exposures per outage day have decreased in the last decade even though the con-

siderable manhours incurred during some of the plant modernisation outages resulted

in increased occupational exposure.

- The average annual exposures in the Asco 1 have been kept on a level comparable to

international values for PWR plants.

Tihange

The plant has been uprated twice since the commissioning. The thermal power of the


The present study focuses on the Tihange 2 second uprate, resulting in a thermal power

level of 110% compared to the initial power level.

- SG replacements have affected outage length but have not significantly affected doses

due to good project organization. Three SG were replaced in 63 days. Typical outage

lengths are between 30 and 40 days.

- Standard PWR water chemistry is maintained, no zinc is injected.

- The exposures during operation are maintained on a constant and rather low level after

the uprate.

- Radiation levels during outage on reactor systems are maintained on a rather low and

constant level after SG Replacement.

- The exposures per outage day have decreased in the last decade even though the con-

siderable manhours incurred during some of the plant modernisation outages resulted

in increased occupational exposure.

- The average annual exposures in the Tihange 2 have been kept on a level comparable

to international values for PWR plants.

Reconstruction experience

Technical factors controlling radiation fields BWR

- BWR plants worldwide continue to show low annual occupational exposure, typically

1 – 2 manSv per plant and year. The lowest exposures experienced are in BWR de-

signs with internal recirculation pumps. No significant difference is seen between

uprated and non-uprated plants.

- A key factor in achieving low radiation levels seems to be maintaining a well-

balanced corrosion product inflow to the primary circuit. Efforts have to be spent to

improve condensate cleanup, especially in connection to power uprates. Zn injection

seems to be an effective method to reduce radiation fields in plants on HWC or

NMCA.

- European BWRs show similar radiation levels as US BWRs in spite of some differ-

ences in reactor design and water chemistry conditions. One Scandinavian BWR dem-

144

onstrates that low radiation fields can be achieved with the combination HWC, Zn in-

jection and ultra-low feedwater Fe (<0.1 ppb).

- In many cases increased steam moisture content is associated with plant power

uprates. Critical reactor components determining the steam moisture content are the

steam separators above the core plus the steam dryer in the top of the reactor pressure

vessel. The lesson learnt is that the design of steam separators and dryer must be care-

fully reviewed when planning an uprate, and that significant redesign of these compo-

nents may be necessary.

- Several plants take actions to reduce the risk of fuel failures of the debris type by in-

troducing fuel with debris filters and by installing cyclotron filtration in the main fe-

edwater lines. The crud induced failures recently seen in some US plants are the result

of bad feedwater chemistry (high levels of Fe, Zn and Cu), and actions are underway

to improve the situation.

Technical factors controlling radiation fields PWR

- A power uprate in a PWR means increased content of boron in the coolant, especially

at the beginning of the cycle. One consequence is a quadratic increase of H-3 produc-

tion. This effect is further enhanced if the plant is operated with long fuel cycles (18 or

24 months).

- Operation with increased Li, a high temperature pH of around 7.4, has shown to have

a favourable impact on radiation levels. Earlier concerns of high-Li operation with re-

spect to fuel material have been reduced due to improved fuel materials. Improved pH

control in most PWRs is the main reason for present rather low radiation fields and

exposures in PWRs. Power uprates with increased boron in the reactor water necessi-

tate further increased reactor water Li in order to maintain the favourable conditions.

One alternative, however costly, is to introduce EBA enriched in B-10 in order to re-

duce the boron content in the coolant.

- An additional way to achieve exposure reduction is to apply Zinc injection with de-

pleted zinc (DZO) in PWRs. The effect of zinc on radiation fields seems to be espe-

cially large in connection to SG replacements.

- One great concern for PWRs is the risk of having fuel problems due to so-called Axial

Offset Anomaly (AOA). The development of AOA is promoted by the combination of

(i) high levels of Ni, e.g. after a SG replacement, (ii) poor pH control with too low Li

levels in BOC, and (iii) high power fuel assemblies with a substantial sub-cooled boil-

ing rate. The combination power uprate and a SG replacement means an increased

risk, and it would be sensible to avoid a power uprate for the first cycle after a SG re-

placement. Factors that reduce the risk is maintaining a good pH control, and operat-

ing in shorter (twelve months) cycles.

- Fuel failure mechanisms in PWRs mostly seem to be associated with fuel design and

materials factors, and only marginally with the power level.

Organizational factors on uprates

145

- Leadership, composition and organisation of the power uprate, including radiological

activities, especially for large demanding tasks are critical for the successful imple-

mentation of power uprate and received doses.

- One important prerequisite for applying the principle of optimisation to occupational

radiation protection is the appropriate and timely exchange of data, techniques and ex-

perience on doses and dose reduction methods.

- It is not recommendable to give the responsibility for important activities to contrac-

tors without strong supervision by plant personnel. Outside personnel are not always

well-trained and acquainted with plant design details as the plant personnel.

- NPPs worldwide have made significant progress over the past 10 years in reducing the

radiation dose to workers. The most important organizational elements in decreasing

exposures are:

� Strong leadership and supervisory monitoring by plant staff

� Recognition of radiation safety as an important responsibility in NPP industry

with its fully integration into the whole project process

� Strict implementation of ALARA principles

� High quality procedures, which have to control all activities which have the

potential for high radiation exposure

� Strict control of entrance into very high radiation areas

� Training and retraining of plant staff and contractors in ALARA activity and

radiation protection

� Operation Experience

� Assessment of Radiation Conditions

- During and after uprates many of the undesirable outcomes can be avoided due to

quality and professional power uprate project organisation, which is of great impor-

tance:

� Project manager and team members should work full time

� Involvement of the operations, radiation safety and training departments early

in the project

� Detailed review process and sufficient time for adequate safety/data analysis,

identification of procedure changes, and reviews during all stages of the pro-

ject, which guarantee that all problems and critical topical area (e.g. fuel) are

properly addressed.

� Sufficient time for training personnel and for incorporating changes in the

control room simulator

� Consideration of existing equipment problems or limitations.

� For extended power uprates it is recommended to use a two-stage implemen-

tation strategy; 1) new fuel design and 2) power uprate

146

� For the Post-Power Uprate Operation phase it should be ensured that the staff

is aware that operating margins are reduced and various plant systems and

components are changed

� Contingency plans should be available, both for the test phase and Post-

Power Uprate Operation, providing guidance to the plant staff when expected

or unexpected conditions are encountered (vibration, flow accelerated corro-

sion, fuel condition, etc)

147

7. References

[1] Turnage K. G,

”Interactions of Fuel and Water Chemistry”,

International Water Chemistry Conference, San Francisco, October 2004

[2] Phung Tran, Joseph Giannelli and Mary Jarvis

”Condensate Polishing Impact on BWR Water Chemistry”,

Int. Conf. on Water Chemistry of Nucl. Reac. Syst., 2006,

Jeju Island, Korea, October 23-26, 2006

[3] Forsberg Stefan, Odelius Helene and Granath Göran,

”Release Rates of Cobalt from Steilite in Simulated Boiling Water Chemistry

Environment”,

International Water Chemistry Conference, San Francisco, October 2004.

[4] Stellwag Bernhard and Staudt Ulrich,

“Water Chemistry Practice at German BWR Plants”,

International Water Chemistry Conference, San Francisco, October 2004

[5] Hettiarachchi S,

”BWR Field Experience with NobleChem”,

Pres. ICONE 8, April 2-6, Baltimore, MD, USA, 2000

[6] Jones Robin L,

“Mitigating Corrosion Problems in LWRs Via Chemistry Changes”,


[7] Robert Cowan, Dennis Hussey,

”Radiation Field Trends as Related to Chemistry in United States BWRs”,

Int. Conf. on Water Chemistry of Nucl. Reac. Syst., 2006,

Jeju Island, Korea, October 23-26, 2006

[8] Johan Lejon,

”Experience of Depleted Zinc Addition to the Oskarshamn 1 and 2 BWRs”,

Presentation at Hamaoka Nuclear Power Plant, Nagoya, Japan 2006

[9] Keys T A,

“Fuel Corrosion Failures in the Browns Ferry Nuclear Plant”,


[10] Bernt Bengtsson,

”RT – Internutbildning Anläggningskompetens – PWR primärkemi”,

RAB Utbildningsmaterial (februari 2004)

[11] ”Pressurized Water Reactor Primary Water Chemistry Guidelines – Revi-

sion 5”, EPRI Final Report 1002884 (September 2003)

[12] PO Aronsson

R2/3/4 - Kompletterande data för Ni och dosrater RAB (oktober 2005)

148

[13] J.M. Partezana et al,

”Survey of Iron and Nickel Concentrations in PWR Primary Coolant”,

EPRI Report 1003136 (August 2001)

[14] Peter Rudling, Tor Ingemansson,

”PWR Fuel Failure Management Handbook – PWR-FFMH”,

Advanced Nuclear Technology International (March 2003)

[15] OECD Nuclear Energy Agency,

”Sixth Annual Report – Occupational exposures at Nuclear Power Plants

1986-1996”,

Information System on Occupational Exposure


”Eighth Annual Report Of the ISOE Programme 1998”,



”Ninth Annual Report – Occupational exposures at Nuclear Power Plants

1999”,



”Eleventh Annual Report – Occupational exposures at Nuclear Power Plants

2001”,Information System on Occupational Exposure


”Twelth Annual Report of the ISOE Programme, 2002 – Occupational expo-

sures at Nuclear Power Plants”,



”Thirteenth Annual Report of the ISOE Programme, 2003 – Occupational

exposures at Nuclear Power Plants”,



”Fourtteenth Annual Report of the ISOE Programme, 2003 – Occupational

exposures at Nuclear Power Plants”,


[22] European Commission,

” Radiation protection 127-Radioactive effluents from nuclear power sta-

tions and nuclear fuel reprocessing plants in the European Union, 1995-

1999”

[23] European Commission,

” Radiation protection 143-Radioactive effluents from nuclear power sta-

tions and nuclear fuel reprocessing plants in the European Union, 1999-

2003”

[24] U.S. Nuclear Regulatory Commission

”Occupational Radiation Exposure at Commercial Nuclear Power Reactors

and Other Facilities 2004- Thrty-Seventh Annual Report”

NUREG-0713, Vol. 26

149

Appendix 1: Data base with NPP uprates

Country 1. Station name 2. Key data 3. Plant type 4. Uprate dataCountry 2.1 Vendor NSSS 2.2 . Comm. Operation 3.1 Reactor type 3.2. Initial power 4.1 Uprate power 4.2 Year implemented 4.3 Uprate type & power increase 4.4 Technical solution 4.5 Equipment

MWt MWt (E, S, MU)* %

Belgium Doel 2 WEC 12.01.75 PWR 1192 1311 2004 E 10% Revision of safety nalyses SGR

Doel 3 FRAM 10.11.82 PWR 2785 3064 1993 E 10% Revision of safety nalyses SGR

Tihange 1 ACLF 09.01.75 PWR 2665 2875 1995 E 8% Revision of safety nalyses SGR

Tihange 2/1995 ACLF 06.01.83 PWR 2785 2905 1995 S 4,3% Use of improved core layout Use of MOX fuel

Tihange 2/2001 ACLF 06.01.83 PWR 2785 3064** 2001 E 10% Revision of safety nalyses SGR

Finland Loviisa 1 AEE 05.09.77 PWR 1375 1500 1998 E 9% Complete revision of all safety analyses

Numerous "tunings" of existing equipment; largest changes in the high-pressure turbine

Loviisa 2 ASEA 01.05.81 PWR 1375 1500 1998 E 9% Complete revision of all safety analyses

Numerous "tunings" of existing equipment; largest changes in the high-pressure turbine

Olkiluoto 1/1984 ASEA 10.01.79 BWR 2000 2160 1984 E 10,8% Revision of safety analyses & substantial

Large modifications in reactor recirculation pumps, turbine,balance of plant equipment (pumps, steam

Olkiluoto 1/1998 ASEA 10.01.79 BWR 2000 2500** 1998** E 25%** turbine, balance of plant equipment (pumps, steam dryer) Modifications in I&C

Olkiluoto 2/1984 ASEA 07.01.82 BWR 2000 2160 1984 E 10,8% Revision of safety analyses & substantial

Large modifications in reactor recirculation pumps

Olkiluoto 2/1998 ASEA 07.01.82 BWR 2000 2500** 1998** E 25%** turbine, balance of plant equipment (pumps, steam dryer)modifications in I&C

Germany Brokdorf KWU 01.11.86 PWR 3765 3900 ? S 3,9%

Emsland KWU 06.20.88 PWR 3765 3850 1990 S 2,3% Increase of average coolant temparature

KWU 06.20.88 PWR 3765 3900 ? S 3,9%

Grafenrheinfeld KWU 06.17.82 PWR 3765 3950 ? S 4,9%

Grohnde KWU 02.01.85 PWR 3765 3850 1990 S 2,3% Increase of average coolant temparature

KWU 02.01.85 PWR 3765 3900 ? S 4,5%

Gundremmingen B KWU 07.01.84 BWR 3840 4100 ? S 6,8%

Gundremmingen C KWU 01.01.85 BWR 3840 4100 ? S 6,8%

Isar-1 KWU 03.21.79 BWR 2575 2755 ? S 7,0%

Isar-2/1991 KWU 04.09.88 PWR 3765 3850 1991 S 2,3% Increase of average coolant temparature

Isar-2/1998 KWU 04.09.88 PWR 3765 3950** 1998** S 4.9%**

Neckar-2/1991 KWU 04.15.89 PWR 3765 3850 1991 S 2,3%

Neckar-2/2005 KWU 04.15.89 PWR 3965** 2005** S 5.3 %**

Philippsburg-2.1991 KWU 12.17.84 PWR 3765 3803 1991 MU 1%

Philippsburg-2/1992 KWU 12.17.84 PWR 3765 3850 1992 S 2.3%

Philippsburg-2/2000 KWU 12.17.84 PWR 3765 3950*** 2000*** S 4.9%*** Increase of thermal reactor power

Unterwester KWU 10.01.79 3733 3900 2000 S 4.5% Increase of thermal reactor power

Hungary PAKS1 AEE 11.03.83 PWR 1375 1500 2006 E 9,1%

PAKS2 AEE 09.21.84 PWR 1375 1500 2006 E 9,1%

PAKS3 AEE 11.03.86 PWR 1375 1500 2007 E 9,1%PAKS4 AEE 11.19.87 PWR 1375 1500 2006 E 9,1%

Japan Higasidory 05.03.05 BWR ? ? ?Shhika HIT 07.30.93 BWR 1593 ? ?

Korea Kori 3 WEC 10.01.86 PWR 2775 2914 2006 S 5%

Kori-4 WEC 04.29.86 PWR 2775 2914 2006 S 5%

YGN 1 WEC 08.25.86 PWR 2775 2914 2006 S 5%

YGN 2 WEC 06.10.87 PWR 2775 2914 2006 S 5%



Mexco LV 1 GE 07.29.90 BWR 1931 2027 1999 S 5%

LV 2 GE 04.10.95 BWR 1931 2027 1999 S 5%

Spain Almaraz-1 WEC 09.01.83 PWR 2696 2739 2003 MU 1.6% Improvement of the feed water flow measurement

Turbine and SG change

Almaraz-2 WEC 07.01.84 PWR 2696 2739 2003 MU 1.6% Improvement of the feed water flow measurement

Turbine and SG change

Asco-1/2000 WEC 12.01.84 PWR 2696 2900 2000 E 8,0% Safety and system Analysis Turbine modification for increased flow

High pressure turbine change Low pressure turbine change Upgrades on main electric equipment

Asco-1/2003 WEC 12.01.84 PWR 2696 2951** 2003 MU 9.5 %** Improvement of the feed water flow measurement

Asco-2/1999 WEC 06.01.85 PWR 2696 2900 1999 E 8,0% Safety and system Analysis Turbine modification for increased flow

Turbine change Upgrades on main electric equipment

Asco-2/2004 WEC 06.01.85 PWR 2696 2952 2004 MU 9.5 %** Improvement of the feed water flow measurement

Cofrentes GE 03.11.85 BWR 2894 2952 1988 S 2% Optimised core layout

Cofrentes GE 03.11.85 BWR 2894 3015 1998 S 4,2% Optimised core layout

Cofrentes GE 03.11.85 BWR 2894 3184 2002 E 10% Fuel loading, turbine optimisation

Cofrentes GE 03.11.85 BWR 2894 3237 2003 E 12% Upgrading of I&C functions

Vandellos-2/1999 WEC 03.08.88 PWR 2775 2900 1999 S 4,5% Safety and system Analysis/No changes on systems except electric

Electric system

Vandellos-2/2002 WEC 03.08.88 PWR 2775 2941 2002 E 10,6% Improvement of the feed water flow measurement

Sweden Forsmark-1 ASEA 12.01.80 BWR 2711 2928 1986 E 8,4% Use of large existing margins Improved fuel

ASEA 12.01.80 BWR 2711 3253 2008 E 19,1%

Forsmark-2 ASEA 07.01.81 BWR 2711 2928 1986 E 8,4% Use of large existing margins Improved fuel

ASEA 07.01.81 BWR 2711 3250 2009 E 19,1%

Forsmark-3 ASEA 08.01.85 BWR 3020 3300 1989 E 9,3% Use of large existing margins Improved fuel

ASEA 08.01.85 BWR 3020 3775 2010 E 25,0%

Oskarshamn-2 ASEA 01.01.75 BWR 1700 1800 1982 S 5,9% Use of large existing margins Improved fuel

Oskarshamn-3 ASEA 08.15.85 BWR 3020 3300 1989 E 9,3% Use of large existing margins Improved fuel

ASEA 08.15.85 BWR 3020 3900 2008 E 29,8%

Ringhals-1 ASEA 01.01.76 BWR 2270 2500 1989 E 10,1% Use of large existing margins Improved fuel

ASEA 01.01.76 BWR 2270 2540 2006 MU 11,9%

Ringhals-2 WEC 05.01.75 PWR 2440 2660 1989 E 9%

SGRRinghals-3 WEC 09.01.81 PWR 2783 3000 2006 E 7,8%

WEC 09.01.81 PWR 2783 3160 2007 S 13,5%

Ringhals-4 WEC 11.21.83 PWR 2783 3160 2011 E 13,5%



Switzerland Leibstadt/1985 GE 12.15.84 BWR 3012 3138 1985 S 4,2% Modification of Turbine System

Leibstadt1998 GE 12.15.84 BWR 3012 3327 1998 S 6%

Leibstadt.1999 GE 12.15.84 BWR 3012 3420 1999 E 9%

Leibstadt/2000 GE 12.15.84 BWR 3012 3515 2000 E 12%

Leibstadt/2003 GE 12.15.84 BWR 3012 3600 2002 E 14,7%

Mühleberg/1976 BBC 11.06.72 BWR 947 997 1976 S 5,3%

Mühleberg/1993 BBC 11.06.72 BWR 947 1047 1993 S 5%

Mühleberg/1994 BBC 11.06.72 BWR 947 1097 1994 E 10%

Gösgen/1985 KWU 11.01.79 PWR 2808 2900 1985 S 3,3%

Gösgen/1992 KWU 11.01.79 PWR 2808 3002 1992 S 6,9%

USA ANO-2 CE 03.26.80 PWR 2815 3026 2002 E 7,5% Revision of Safety analyses (SLB increase design containment pressure), revision of transition reload safety analyses

SGR, modifications to the BOP changes to the Main Unit Turbine/Generator, the Main Unit Condenser, and accessories and associated supporting systems.

Beaver Valley 1/2001 WEC 10.01.76 PWR 2652 2689 2001 MU 1,4%

Beaver Valley 1/2006 WEC 10.01.76 PWR 2652 2900 2006 E 9,4% Revision of Safety analyses Upgrading of existing eqquipment, component i

Beaver Valley 2/2001 WEC 11.17.87 PWR 2652 2689 2001 MU 1,4%

Beaver Valley 2/2006 WEC 11.17.87 PWR 2652 2900 2006 E 9,4% Revision of Safety analyses Upgrading of existing eqquipment, component i

Braidwood 1 WEC 07.29.88 PWR 3411 3587 2001 S 5,2%

Braidwood 2 WEC 10.17.88 PWR 3411 3587 2001 S 5,2%

Browns Ferry 2/1998 GE 03.01.75 BWR 3294 3458 1998 S 5,0%

Browns Ferry 2/2007 GE 03.01.75 BWR 3294 3952 2007 E 20% Revision of Safety analyses, updated analysis methods to predict the load's experience by its steam dryers during power uprate operation.

Upgrading of existing eqquipment, component instrumentation of a steam dryer to collect actual pressure load data during extended power uprate operation,

Browns Ferry 3 GE 03.01.75 BWR 3294 3458 1998 S 5,0%

Browns Ferry 3/2007 GE 03.01.75 BWR 3294 3952 2007 E 20% Revision of Safety analyses, updated analysis methods to predict the load's experience by its steam dryers during power uprate operation.


Browns Ferry 1 GE 01.08.74 BWR 3293 3952 2007 E 20% Revision of Safety analyses, updated analysis methods to predict the load's experience by its steam dryers during power uprate operation.


Brunswick 1/1996 GE 03.18.77 BWR 2436 2558 1996 S 5,0%

Brunswick 1/2002 GE 03.18.77 BWR 2436 2923 2002 E 20,0%

Brunswick 2/1996 GE 11.03.75 BWR 2436 2558 1996 S 5,0%



USA Brunswick 2/2002 GE 11.03.75 BWR 2436 2923 2002 E 20,0% increasing steam production, holding liquid flow in the core, dome pressure and temperatures near current values. The increased steam production is achieved by "flattening" the core power profile, which involves increasing power generation in the outer regions of the core. There is an increase in feedwater flow to match the increased production of steam. performed in two steps

BOP modification

Byron 1 WEC 09.16.85 PWR 3411 3587 2001 S 5,2%

Byron 2 WEC 08.21.87 PWR 3411 3587 2001 S 5,2%

Callaway WEC 04.09.85 PWR 3411 3565 1988 S 4,5%

Calvert Cliffs 1/1977 CE 05.08.75 PWR 2560 2700 1977 S 5,5%Calvert Cliffs 1/2006 CE 05.08.75 PWR 2560 2737 2006 MU 6,8%Calvert Cliffs 2/1977 CE 04.01.77 PWR 2560 2700 1977 S 5,5%

Calvert Cliffs 2/2006 CE 04.01.77 PWR 2560 2700 2006 MU 6,8%

Clinton GE 11.24.87 BWR 2894 3473 2002 E 20,0% Revision of the design basis accident (higher steam and feedwater flows). The constant-pressure power uprate Done in two steps of 7 and 13%. the unit is operating within the power range of other BWR/6 nuclear steam supply systems.

New design of the fuel, BOP modification

Comanche Peak 1 WEC 08.13.90 PWR 3411 3458 2001 MU 1,4%

Comanche Peak 2 WEC 08.03.93 PWR 3411 3445 1999 MU 1,0%Comanche Peak 2 3458 2001 MU 1,4%Crystal River 3 BW 03.13.77 PWR 2544 2568 2002 S 0,9%

D.C. Cook 1 WEC 05.07.85 PWR 3250 3304 2002 MU 1,7%

D.C. Cook 2 WEC 03.13.86 PWR 3411 3468 2003 MU 1,7%Diablo Canyon 1 WEC 05.07.85 PWR 3338 3411 2000 S 2,2%

Dresden 2 GE 06.09.70 BWR 2527 2957 2001 E 17,0% Increase of main steam and feedwater flows

modification of high-pressure turbine

Dresden 3 GE 11.16.71 BWR 2527 2957 2001 E 17,0% Increase of main steam and feedwater flows


Duane Arnold/1985 GE 02.01.75 BWR 1593 1658 1985 S 4,1%

Duane Arnold/2001 GE 02.01.75 BWR 1593 1912 2001 E 19,4% increasing steam production, holding liquid flow in the core, dome pressure and temperatures near current values. The increased steam production is achieved by "flattening" the core power profile, which involves increasing power generation in the outer regions of the core. There is an increase in feedwater flow to match the increased production of steam. performed in two steps

BOP modification

Farley 1 WEC 12.01.77 PWR 2652 2775 1998 S 4.6%

Farley 2 WEC 07.30.81 PWR 2652 2775 1998 S 4,6%

Fermi 2 GE 01.23.88 BWR 3293 3430 1992 S 4,2%

Fitzpatrick GE 07.28.75 BWR 2436 2536 1996 S 4,1%

Fort Calhoun/1980 CE 09.26.73 PWR 1420 1500 1980 S 5,6%Fort Calhoun/2004 CE 09.26.73 PWR 1420 1525 2004 MU 7,4%

Fort Calhoun/2006 CE 09.26.73 PWR 1420 1547 2006 MU 8,9%

Ginna WEC 07.01.70 PWR 1520 1775 2006 E 16.8%



USA Grand Gulf GE 07.01.85 BWR 3833 3898 2002 MU 1,7%

H. B. Robinson/1979 WEC 03.07.71 PWR 2200 2300 1979 S 4,5%H. B. Robinson/2002 WEC 03.07.71 PWR 2200 2339 2002 MU 6,3%

Hatch 1/1995 GE 01.01.76 BWR 2436 2558 1995 S 5,0%

Hatch 1/1998 GE 01.01.76 BWR 2436 2763 1998 E 13,4%

Hatch 1/2003 GE 01.01.76 BWR 2436 2804 2003 MU 15,1%

Hatch 2/1995 GE 09.05.79 BWR 2436 2558 1995 S 5,0%

Hatch 2/1998 GE 09.05.79 BWR 2436 2763 1998 E 13,4%

Hatch 2/2003 GE 09.05.79 BWR 2436 2804 2003 MU 15,1%

Hope Creek GE 12.20.86 BWR 3293 3339 2001 MU 1,4%

Indian Point 2/2003 WEC 08.01.74 PWR 3071 3114 2003 MU 1,4%

Indian Point 2/2004 WEC 08.01.74 PWR 3071 3216 2004 S 4.7%

Indian Point 3/2002 WEC 08.30.76 PWR 3025 3067 2002 MU 1,4%

Indian Point 3/2005 WEC 08.30.76 PWR 3025 3216 2005 S 6.3%

Kewanee /2003 WEC 06.16.74 PWR 1650 1673 2003 MU 1,4%

Kewanee /2004 WEC 06.16.74 PWR 1650 1772 2004 S 7.4%

LaSalle 1 GE 01.01.84 BWR 3323 3489 2000 S 5,0% Increase of main steam and feedwater flows

modification of BOP

LaSalle 2 GE 01.01.85 BWR 3323 3489 2000 S 5,0% Increase of main steam and feedwater flows

modification of BOP

Limerick 1 GE 02.01.86 BWR 3293 3458 1996 S 5,0% Increase of main steam and feedwater flows

modification of BOP

Limerick 2 GE 01.08.90 BWR 3293 3458 1995 S 5,0% Increase of main steam and feedwater flows

modification of BOP

Millstone 2 CE 12.26.75 PWR 2560 2700 1979 S 5,5%

Monticello GE 06.30.71 BWR 1670 1775 1998 E 6,3% Increase of main steam and feedwater flows

modification of BOP, new design of fuel

Nine Mile Point 2 GE 03.11.88 BWR 3323 3467 1995 S 4,3% Increase of main steam and feedwater flows

modification of BOP, new design of fuel

North Anna 1 WEC 06.06.78 PWR 2775 2893 1986 S 4,3%

North Anna 2 WEC 12.14.80 PWR 2775 2893 1986 S 4,3%

Palisades CE 12.31.71 PWR 2530 2565 2004 MU 1,4%

Palo Verde 1/1996 CE 02.13.86 PWR 3800 3876 1996 S 2,0%



Palo Verde 2/2003 CE 09.19.86 PWR 3990 2003 S 5,0%



Peach Bottom 2/1994 GE 07.05.74 BWR 3293 3458 1994 S 5,0%

Peach Bottom 2/2002 GE 07.05.74 BWR 3293 3514 2002 MU 6,7%

Peach Bottom 3/1995 GE 12.23.74 BWR 3293 3458 1995 S 5,0%

Peach Bottom 3/2002 GE 12.23.74 BWR 3293 3514 2002 MU 6,7%

Perry GE 11.18.87 BWR 3579 3758 2000 S 5,0%

Pilgrim GE 12.09.72 BWR 1998 2028 2003 MU 1,5%

Point Beach 1 WEC 12.21.70 PWR 1519 1540 2002 MU 1,4%

Point Beach 2 WEC 09.30.72 PWR 1519 1540 2002 MU 1,4%



USA Quad Cities 1 GE 02.18.73 BWR 2511 2957 2001 E 17,8% Increase of main steam and feedwater flows


Quad Cities 2 GE 03.10.73 BWR 2511 2957 2001 E 17,8% Increase of main steam and feedwater flows


River Bend/2000 GE 06.16.86 BWR 2894 3039 2000 S 5,0% two phases: 1.steam flow/feedwater flow increase in power (only flow) 2.flow increase/reactor pressure increase

modification of high-pressure turbine, new fuel design

River Bend/2003 GE 06.16.86 BWR 2894 3091 2003 MU 6,7%

Salem 1/1986 WEC 06.30.77 PWR 3344 3411 1986 S 2,0%

Salem 1/2001 WEC 06.30.77 PWR 3344 3459 2001 MU 3,4%

Salem 2 WEC 10.13.81 PWR 3411 3459 2001 MU 1,4%

San Onofre 2 CE 08.08.83 PWR 3390 3438 2001 MU 1,4%

San Onofre 3 CE 04.01.84 PWR 3390 3438 2001 MU 1,4%

Seabrook1/2005 WEC 08.19.90 PWR 3411 3587 2005 S 5.2%Seabrook1/2006 WEC 08.19.90 PWR 3411 3648 2006 S 6.9%Sequoyah 1 WEC 07.01.81 PWR 3411 3455 2002 MU 1,3%

Sequoyah 2 WEC 06.01.82 PWR 3411 3455 2002 MU 1,3%

Shearon Harris WEC 05.02.87 PWR 2775 2900 2001 S 4,5%

South Texas 1 WEC 08.25.88 PWR 3800 3853 2002 MU 1,4%

South Texas 2 WEC 06.18.89 PWR 3800 3853 2002 MU 1,4%

St. Lucie 1 CE 12.21.76 PWR 2560 2700 1981 S 5,5%

St. Lucie 2 CE 08.08.83 PWR 2560 2700 1985 S 5,5%

Surry 1 WEC 12.02.72 PWR 2441 2546 1995 S 4,3%

Surry 2 WEC 05.01.73 PWR 2441 2546 1995 S 4,3%

Susquehanna 1/1995 GE 06.08.83 BWR 3293 3441 1995 S 4,5%

Susquehanna 1/2001 GE 06.08.83 BWR 3293 3489 2001 MU 6,0%

Susquehanna 2/1994 GE 02.12.85 BWR 3293 3441 1994 S 4,5%

Susquehanna 2/12001 GE 02.12.85 BWR 3293 3489 2001 MU 6,0%

TMI-1 BW 09.02.74 PWR 2535 2568 1988 S 1,3%

Turkey Point 3 WEC 12.04.72 PWR 2200 2300 1996 S 4,5%

Turkey Point 4 WEC 09.07.73 PWR 2200 2300 1996 S 4,5%

Vermont Yankee GE 11.29.72 BWR 1593 1912 2006 E 20.0% Revision of Safety analyses, updated analysis methods to predict the load's experience by its steam dryers during power uprate operation.


V. C. Summer WEC 12.02.72 PWR 2775 2900 1996 S 4,5%

Vogtle 1 WEC 05.31.87 PWR 3411 3565 1993 S 4,5%

Vogtle 2 WEC 05.19.89 PWR 3411 3565 1993 S 4,5%

Waterford 3/2002 CE 09.24.85 PWR 3390 3441 2002 MU 1,5%

Waterford 3/2005 CE 09.24.85 PWR 3390 3665 2005 E 8,0%

Watts Bar WEC 05.27.96 PWR 3411 3459 2001 MU 1,4%

WNP-2 BWR 3323 3486 1995 S 4,9%



USA Wolf Creek WEC 09.03.85 PWR 3411 3565 1993 S 4,5%

Slovenia Krsko WEC 01.01.83 PWR 1882 2000 2000 S 6,3% Use additional margin by new SG-higher heat transfer area

SGR

*S- Stretch power (uprates typically up to 7 %, within the design capacity plant). Typically involve changes to instrumentation setpoints but do not involve major plant modifications)MU-Measurement uncertainty (uprates are less than 2 %e achieved by implementing enhanced techniques for calculating reactor power) E-Extended power (uprates greater than stretch. Require significant modifications to major BOP equipment) **Second thermal uprate***Third thermal uprate

PWRBWR

Country 1. Station name 2. Key data 3. Plant typeCountry 2.1 Vendor NSSS 2.2 . Comm. Operation 3.1 Reactor type 3.2. Initial power

MWt

Belgium Doel 2 WEC 12.01.75 PWR 1192

Doel 3 FRAM 10.11.82 PWR 2785

Tihange 1 ACLF 09.01.75 PWR 2665

Tihange 2/1995 ACLF 06.01.83 PWR 2785

Tihange 2/2001 ACLF 06.01.83 PWR 2785

Finland Loviisa 1 AEE 05.09.77 PWR 1375

Loviisa 2 ASEA 01.05.81 PWR 1375

Olkiluoto 1/1984 ASEA 10.01.79 BWR 2000




Germany Brokdorf KWU 01.11.86 PWR 3765

Emsland KWU 06.20.88 PWR 3765

KWU 06.20.88 PWR 3765

Grafenrheinfeld KWU 06.17.82 PWR 3765

Grohnde KWU 02.01.85 PWR 3765

KWU 02.01.85 PWR 3765

Gundremmingen B KWU 07.01.84 BWR 3840

Gundremmingen C KWU 01.01.85 BWR 3840

Isar-1 KWU 03.21.79 BWR 2575

Isar-2/1991 KWU 04.09.88 PWR 3765

Isar-2/1998 KWU 04.09.88 PWR 3765

Neckar-2/1991 KWU 04.15.89 PWR 3765

Neckar-2/2005 KWU 04.15.89 PWR

Philippsburg-2.1991 KWU 12.17.84 PWR 3765

Philippsburg-2/1992 KWU 12.17.84 PWR 3765


Unterwester KWU 10.01.79 3733

Hungary PAKS1 AEE 11.03.83 PWR 1375

PAKS2 AEE 09.21.84 PWR 1375

PAKS3 AEE 11.03.86 PWR 1375PAKS4 AEE 11.19.87 PWR 1375

Japan Higasidory 05.03.05 BWR ?Shhika HIT 07.30.93 BWR 1593

Korea Kori 3 WEC 10.01.86 PWR 2775

Kori-4 WEC 04.29.86 PWR 2775

YGN 1 WEC 08.25.86 PWR 2775

YGN 2 WEC 06.10.87 PWR 2775

5. Fuel 6. Annual liquid effluent releases 7. Annual gases effluent releases 5.1 Fuel cycle 5.2 Average linear

fuel rating BU5.3 Average linear fuel rating AU

5.4 Fuel type 6.1 Before the uprate 6.2 After the uprate 7.1 Before the uprate 7.2 After the uprate

months kW/m kW/m GBq GBq GBq GBq

11 22.22 22.22 FRA-AFA/KWU-AKA 4803 tritium 99,9% year 2003 (14% of total for 4 units)

208 tritium 52% year 2003 (14% of total for 4units)

11 20 20 KWU-AKA/FRA-AFA/ABB-PAAD 15805 tritium 99.9% year 1992 (36% of total for 4 units)

11811 tritium 99.9% year 1994 (36% of total for 4 units)


1066 tritium 67% year 1994 (36% of total for 4 units)

18 22.15 22.15 FRA Std/W Std/Exxon Std 11041 tritium 99.9 % year 1994 (1/3 of total for 3 units)

14917 tritium 99.9 % year 1996 (1/3 of total for 3 units)

5617 tritium 29 % year 1994 (1/3 of total for 3 units)


15 17.85 17.85 FRA Std/ FRA AFA 11041 tritium 99.9 % year 1994 (1/3 of total for 3 units)

14917 tritium 99.9 % year 1996 (1/3 of total for 3 units)



15 17.85 17.85 FRA Std/ FRA AFA 11033 tritium 99.9% year 2000 (1/3 of total for 3 units)

19867 tritium99.9% year 2002 (1/3 of total for 3 units)

3693 tritium 68% year 2000 (1/3 of total for 3 units)


12 32.5* * peak linear fuel rating

32.5* * peak linear fuel rating

6000 tritium 99.9% year 1997 (1/2 of total for 2 units)




12 32.5* * peak linear fuel rating






12 17.90 4 MW/Assembly

17.90 4.32 MW/Assembly

8x8-1

12 17.90 4.32 MW/Assembly

14.90 5 MW/Assembly

9x9-1/Altrium 10B "Novel BWR fuel"







8x8-1


13.10 5 MW/Assembly

SVEA-100/GE12 "Novel BWR fuel"





12 16.67 16.67 KWU Convoy 8300 tritium 99.9% year 1991

780 tritium 86% year 1991

12 21.10 21.10 16000 tritium 99.9% year 1991


12 16.70 KWU 7200 tritium 99.9% year 1990

16000 tritium 99.9% year 1992



12 17.10 WEST/KWU 1140 tritium 85% year 1997


12 16740 tritium 99.9% year 1990 (62% of total for 2 units)




12

12 23.14 KWU 16x16-20 19000 tritium 99.9% year 1990

15001 tritium 99.9% year 1992



12 23.14 KWU 16x16-20 17000 tritium 99.9% year 1991

13000 tritium 99.9% year 1993



12 29.70 KWU/Fragema 18000 tritium 99.9% year 1999

13000 tritium99.9% year 2001



11 20.50 20.50 7700 tritium 99.9% year 1999

16000 tritium99.9% year 2001



15 14 8x8 high burnup

18 17.83 V-5H

18 17.83 V-5H

18 17.83 V-5H

18 17.83 V-5H


MWt

Mexco LV 1 GE 07.29.90 BWR 1931

LV 2 GE 04.10.95 BWR 1931

Spain Almaraz-1 WEC 09.01.83 PWR 2696

Almaraz-2 WEC 07.01.84 PWR 2696

Asco-1/2000 WEC 12.01.84 PWR 2696

Asco-1/2003 WEC 12.01.84 PWR 2696

Asco-2/1999 WEC 06.01.85 PWR 2696

Asco-2/2004 WEC 06.01.85 PWR 2696

Cofrentes GE 03.11.85 BWR 2894




Vandellos-2/1999 WEC 03.08.88 PWR 2775


Sweden Forsmark-1 ASEA 12.01.80 BWR 2711

ASEA 12.01.80 BWR 2711

Forsmark-2 ASEA 07.01.81 BWR 2711

ASEA 07.01.81 BWR 2711


ASEA 08.01.85 BWR 3020

Oskarshamn-2 ASEA 01.01.75 BWR 1700


ASEA 08.15.85 BWR 3020

Ringhals-1 ASEA 01.01.76 BWR 2270

ASEA 01.01.76 BWR 2270

Ringhals-2 WEC 05.01.75 PWR 2440


WEC 09.01.81 PWR 2783










18 17.30 17.30 "4.35" 9148 tritium 96% year 2002 (1/2 of total for 2 units)

18 17.30 17.30 "4.35" 9148 tritium 96% year 2002 (1/2 of total for 2 units)

12 17.30 3.5 avg enrich

18.92 4.7 avg enrich

STD/ MAEF-IFM (Enusa)





12 18.92 MAEF-IFM (Enusa) 2655 tritium 56% year 2002 (1/2 of total for 2 units)

12 17.30 3.5 avg enrich

18.92 4.7 avg enrich

STD/ MAEF-IFM (Enusa)





12 18.92 4.7 avg enrich

MAEF-IFM (Enusa) 5251 tritium 19% year 2003 (1/2 of total for 2 units)

18 4.64 MW/Assembly 4.73 MW/Assembly Vale

18 4.73 MW/Assembly 20.03 4.83 MW/Assembly

GE10/GE11 511 tritium 99.9% year 1997

227 tritium 99.9% year 1999





SVEA96/GE12/GE14/Altrium 1250 tritium 99.9% year 2001

1020 tritium 99.9% year 2003





SVEA96/GE12/GE14/Altrium 21036 tritium 19% year 2002

18 17.80 3.6 avg enric

OFA/ AEF 33509 tritium 99.9% year 1998

35727 tritium 99.9% year 2000



18 18.92 4.7 avg enrich

MAEF-IFM (Enusa) 10818 tritium 99.9% year 2001

32325 tritium 99.9% year 2003



12 4.01 MW/Assembly

106 ?? 4.33 MW/Assembly

8x8/9x9 759 tritium 76% year 1985 (1/2 of total for 2 units)


71107 tritium insignificant, year 1985


12 4.01 MW/Assembly

106 ?? 4.33 MW/Assembly

8x8/10x10 759 tritium 76% year 1985 (1/2 of total for 2 units)




12 4.31 MW/Assembly


8x8/Altrium 10B 395 tritium 99% year 1988




12 3.83 MW/Assembly

39* peak linear fuel rating 4.05 MW/Assembly

8x8/Altrium 10 361 tritium 83% year 1981 (1/2 of total for 2 units)




12 4.31 MW/Assembly

41.5* peak linear fuel rating 4.71 MW/Assembly

8x8/10x10 438 tritium 98% year 1988




11 3.50 MW/Assembly

40 28 3.85 MW/Assembly

ABB 8 x8/ ABB SUEA64/SUEA100/ Altrium10B





12

22.02

W/ KWU/W FRA-ANP AFA 36


69760 tritium not measurd year 1990 (32% of total for 3 units)


MWt

Switzerland Leibstadt/1985 GE 12.15.84 BWR 3012

Leibstadt1998 GE 12.15.84 BWR 3012

Leibstadt.1999 GE 12.15.84 BWR 3012

Leibstadt/2000 GE 12.15.84 BWR 3012


Mühleberg/1976 BBC 11.06.72 BWR 947



Gösgen/1985 KWU 11.01.79 PWR 2808

Gösgen/1992 KWU 11.01.79 PWR 2808

USA ANO-2 CE 03.26.80 PWR 2815

Beaver Valley 1/2001 WEC 10.01.76 PWR 2652




Braidwood 1 WEC 07.29.88 PWR 3411


Browns Ferry 2/1998 GE 03.01.75 BWR 3294


Browns Ferry 3 GE 03.01.75 BWR 3294



Brunswick 1/1996 GE 03.18.77 BWR 2436

Brunswick 1/2002 GE 03.18.77 BWR 2436

Brunswick 2/1996 GE 11.03.75 BWR 2436






GE6



GEA 6,7,8/SVEA 96



GEA 6,7,8/SVEA 96



GEA 6,7,8/SVEA 96



GEA 6,7,8,10/SVEA 96

12 3.95 MW/Assembly 4.15 MW/Assembly GE


GE 11

12 4.36 MW/Assembly 39.70/ 14.6 4.57 MW/Assembly

GE 11/ GE14

12 RBU

12 22.6 12000 tritium 99.9% year 1991

13000 tritium 99.9% year 1993

5100 tritium not measurd year 1991

11000 tritium not measurd year 1993

18 18.0

18 17.06 17.06 West Vantage 5 22904 tritium 99,99%, year2000

13124 tritium 99.99%, year 2002

11359 tritium 84,5%, year 2000

5907 tritium 88.0%, year 2002

18 17.06 17.06 West Vantage 5 1205 tritium 80.0%, year 2002

18 18.3 18.3 49103 tritium 99,99%, year2000

43327 tritium 99,99%, year 2002

656 tritium 97,5%, year 2000

176 tritium 91,2%, year 2002

18 18.3 18.3 51487 tritium 99,99%, year 2000

43327 tritium 99,99%, year 2002

698 tritium 97,1%, year 2000

36 tritium 26,2%, year 2002

24 18.49 4.31 MW(Assembly


3E 98



3E 98



GE 98



GE 98



GE 7B

24 4.35 MW/Assembly

18.42 4,56 MW/Assembly

24 18.42 4,56 MW/Assembly


4514 tritium 99,99%, year 2001

2234 tritium 99,99%, year 2003

27861 tritium 19,4%, year 2001

9805 tritium 35,7%, year 2003

24 4.35 MW/Assembly

18.61 4,56 MW/Assembly


MWt

USA Brunswick 2/2002 GE 11.03.75 BWR 2436

Byron 1 WEC 09.16.85 PWR 3411

Byron 2 WEC 08.21.87 PWR 3411

Callaway WEC 04.09.85 PWR 3411

Calvert Cliffs 1/1977 CE 05.08.75 PWR 2560Calvert Cliffs 1/2006 CE 05.08.75 PWR 2560Calvert Cliffs 2/1977 CE 04.01.77 PWR 2560

Calvert Cliffs 2/2006 CE 04.01.77 PWR 2560

Clinton GE 11.24.87 BWR 2894

Comanche Peak 1 WEC 08.13.90 PWR 3411

Comanche Peak 2 WEC 08.03.93 PWR 3411Comanche Peak 2Crystal River 3 BW 03.13.77 PWR 2544

D.C. Cook 1 WEC 05.07.85 PWR 3250

D.C. Cook 2 WEC 03.13.86 PWR 3411Diablo Canyon 1 WEC 05.07.85 PWR 3338

Dresden 2 GE 06.09.70 BWR 2527


Duane Arnold/1985 GE 02.01.75 BWR 1593


Farley 1 WEC 12.01.77 PWR 2652


Fermi 2 GE 01.23.88 BWR 3293

Fitzpatrick GE 07.28.75 BWR 2436

Fort Calhoun/1980 CE 09.26.73 PWR 1420Fort Calhoun/2004 CE 09.26.73 PWR 1420

Fort Calhoun/2006 CE 09.26.73 PWR 1420

Ginna WEC 07.01.70 PWR 1520





24 18.61 4,56 MW/Assembly


18 18.30 18.30 42809 tritium 99,9%, year2000

35150 tritium 99,9%, year 2002

74 tritium 40,7%, year 2000

148 tritium 57,4%, year 2002

18 18.30 18.30 42809 tritium 99,9%, year 2000

35113 tritium 99,9%, year 2002

125 tritium 66,5%, year 2000

131 tritium 80,8%, year 2002

18 19.13 19.13 Vantage 5

24 20.62 20.62 C-E 14x14

24 20.56 20.56 C-E 14x14



GE6/GE7B/GE8B insignificant, year 2000 insignificant, year 2004 1554 tritium 99,9%, year 2000

2072 tritium 57,3%, year 2004

18 17.81 17.81 SIEMENS 5609 tritium 99.9%, year 2000

51258 tritium 99.9%, year 2002

1133 tritium 96,6%, year 2000

10541 tritium 20.0%, year 2002

18 17.8 17.8 SIEMENS18 17.8 17.8 SIEMENS24 18.67 18.67 MARK B4Z 12366 tritium 99.9%, year

200125930 tritium 99.9%, year 2003

16078 tritium 3.3%, year 2001

6317 tritium 4.3%, year 2003

18 21.98 21.98 insignificant, year 2001 insignificant, year 2003 4908 tritium 85.0%, year 2001

5341 tritium 93.0%, year 2003

18 17.81 17.8118 17.50 17.50 West Vantage 5 40515 tritium 99.9%,

year 20019602 tritium 85.0%, year 2001

24 47.6* peak linear fuelrating 3.49 MW/Assembly


24 47.6* peak linear fuelrating 3.49 MW/Assembly


24 4.33 MW/Assembly 4.51 MW/Assembly GE 8x8 barrier



18 17900 tritium 99.9% year 1997 (1/2 of total for 2 units)


18 17900 tritium 99.9% year 1997 (1/2 of total for 2 units)




24 4.35 MW/Assembly

4.53 MW/Assembly

18 Standard18 49.93*

* peak linear fuel rating


10900 tritium 99.9%, year 2003



MWt

USA Grand Gulf GE 07.01.85 BWR 3833

H. B. Robinson/1979 WEC 03.07.71 PWR 2200H. B. Robinson/2002 WEC 03.07.71 PWR 2200

Hatch 1/1995 GE 01.01.76 BWR 2436

Hatch 1/1998 GE 01.01.76 BWR 2436

Hatch 1/2003 GE 01.01.76 BWR 2436

Hatch 2/1995 GE 09.05.79 BWR 2436

Hatch 2/1998 GE 09.05.79 BWR 2436

Hatch 2/2003 GE 09.05.79 BWR 2436

Hope Creek GE 12.20.86 BWR 3293

Indian Point 2/2003 WEC 08.01.74 PWR 3071




Kewanee /2003 WEC 06.16.74 PWR 1650

Kewanee /2004 WEC 06.16.74 PWR 1650

LaSalle 1 GE 01.01.84 BWR 3323


Limerick 1 GE 02.01.86 BWR 3293


Millstone 2 CE 12.26.75 PWR 2560

Monticello GE 06.30.71 BWR 1670

Nine Mile Point 2 GE 03.11.88 BWR 3323

North Anna 1 WEC 06.06.78 PWR 2775


Palisades CE 12.31.71 PWR 2530

Palo Verde 1/1996 CE 02.13.86 PWR 3800

Palo Verde 1/2005 CE 02.13.86 PWR 3800

Palo Verde 2/1996 CE 09.19.86 PWR 3800

Palo Verde 2/2003 CE 09.19.86 PWR

Palo Verde 3/1996 CE 01.15.88 PWR 3800

Palo Verde 3/2005 CE 01.15.88 PWR 3800

Peach Bottom 2/1994 GE 07.05.74 BWR 3293




Perry GE 11.18.87 BWR 3579

Pilgrim GE 12.09.72 BWR 1998

Point Beach 1 WEC 12.21.70 PWR 1519








ANF 8x8/GE 8x8R 2238 tritium 99.9%, year 2001

861 tritium 99.9%, year 2003

4551 tritium 62,4%, year 2001

3639 tritium 62.1%, year 2003

18 ANF Standard18 43.96*

* peak linear fuel rating


12443 tritium 99.9% year 2001

6127 tritium 99.9% year 2003



24 4.35 MW/Assembly 4.56 MW/Assembly


24 4.93 MW/Assembly 5.00 MW/Assembly 593 tritium 99.9%, year 2002

463 tritium 99.9%, year 2004

1166 tritium 44.2%, year 2002

1760 tritium 35.5%, year 2004



24 4.93 MW/Assembly 5.00 MW/Assembly 351 tritium 99.9%, year 2002

741 tritium 99.9%, year 2004

1242 tritium 49.3%, year 2002

2168 tritium 41.3%, year2004



GE7/GE9 barrier 144 tritium 99.9%, year 2002

819 tritium 59.3%, year 2002

24 18.81 18.81

24 21 21 OFA/V-5 34258 tritium 99.9% year 2001

35309 tritium 99.9% year 2003



18 20.82 20.82 3327 tritium 99.9% year 2002

17368 tritium 99.9% year 2004



10360 tritium 99.9% year 2003



44 4.57 MW/Assembly

GE6/Atrium 9B insignificant, year 2001 101040 tritium 9%, year2001


44 4.57 MW/Assembly

GE6/Atrium 9B

24 16.40 4.31MW/Assembly

16.40 4.53MW/Assembly

GE7B/GE8B/GE9B/GE11



GE7B/GE9B/GE11

18 18.5 18.5


39 3.67 MW/Assembly

8x8/8x8R/P8x8R/BP8x8



GE6B/GE11

18 18.59 18.59 West Standard

18 18.59 18.59 West Standard

18 25.73 25.73 Siemens 7310 tritium 99.9%, year 2003

3227 tritium 25.5%, year 2003

18 18.14 18.14 C-E System 80 no data no data 18633 tritium 78%, year 1995 (1/3 of total for 3 units)

no data

no data 25726 tritium 96.7%, year 2004

18 no data no data 18633 tritium 78%, year 1995 (1/3 of total for 3 units)

no data

18.21 18.21 C-E System 80 no data no data 16098 tritium 62%, year 2002

49231 tritium 99%, year 2004

18 18.37 18.37 C-E System 80 no data no data 18633 tritium 78%, year 1995 (1/3 of total for 3 units)

no data

no data 36881 tritium 98%, year 2004

24 4.31MW/Assembly

4.53MW/Assembly



GE8B/GE9B/GE11 1308 tritium 99.9%, year 2001

609 tritium 99.9%, year 2003


29726 tritium 2.4%, year2003

24 4.31MW/Assembly

4.53MW/Assembly



GE7B/GE8B/GE9B

15-18 19.85 4.78 MW/Assembly


no data for year 1999 1330 tritium 99.9%, year 2001

no data for year 1999 2093 tritium insignificant, year 2001

24 44* peak linear fuel rating 3.44 MW/Assembly

44* peak linear fuel rating 3.50 MW/Assembly

insignificant, year 2002 no data for year 2004 27842 tritium 92.0%, year 2002

12427 tritium 84.0%, year 2004

18 18.7 18.7 OFA 21331 tritium 99.9% year 2001

27683 tritium 99.9% year 2003

3007 tritium 98.0% year 2001

2276 tritium 99.9% year 2003

18 18.7 18.7 OFA


MWt

USA Quad Cities 1 GE 02.18.73 BWR 2511

Quad Cities 2 GE 03.10.73 BWR 2511

River Bend/2000 GE 06.16.86 BWR 2894

River Bend/2003 GE 06.16.86 BWR 2894

Salem 1/1986 WEC 06.30.77 PWR 3344

Salem 1/2001 WEC 06.30.77 PWR 3344

Salem 2 WEC 10.13.81 PWR 3411

San Onofre 2 CE 08.08.83 PWR 3390


Seabrook1/2005 WEC 08.19.90 PWR 3411Seabrook1/2006 WEC 08.19.90 PWR 3411Sequoyah 1 WEC 07.01.81 PWR 3411

Sequoyah 2 WEC 06.01.82 PWR 3411

Shearon Harris WEC 05.02.87 PWR 2775

South Texas 1 WEC 08.25.88 PWR 3800


St. Lucie 1 CE 12.21.76 PWR 2560

St. Lucie 2 CE 08.08.83 PWR 2560

Surry 1 WEC 12.02.72 PWR 2441

Surry 2 WEC 05.01.73 PWR 2441

Susquehanna 1/1995 GE 06.08.83 BWR 3293




TMI-1 BW 09.02.74 PWR 2535

Turkey Point 3 WEC 12.04.72 PWR 2200


Vermont Yankee GE 11.29.72 BWR 1593

V. C. Summer WEC 12.02.72 PWR 2775

Vogtle 1 WEC 05.31.87 PWR 3411


Waterford 3/2002 CE 09.24.85 PWR 3390

Waterford 3/2005 CE 09.24.85 PWR 3390

Watts Bar WEC 05.27.96 PWR 3411

WNP-2 BWR 3323













GE/Siemens Power Corporation fuel

478 tritium 99.9%, year 1999

1721 tritium 99.9%, year 2001

29488 tritium 1.0%, year 1999

1237 tritium 34.0%, year 2001



GE7B/GE8B 3404 tritium 99.9%, year 2002

3944 tritium 99.9%, year 2002

1581 tritium 35.4%, year 2002

2977 tritium 45.2%, year 2004

18 West Standard

18 17.85 17.85 26344 tritium 99.9% year 2002

14116 tritium 52.2% year 2002

18 17.85 17.85 West Standard 8166 tritium 99.9% year 2002

39564 tritium 18.4% year 2002

18 18.4 18.4 55078 tritium 99.9% year 2002

3989 tritium 41.0% year 2002

18 18.4 18.4

18

18 17.85 17.85 West Vantage 5H 51164 tritium 99.9% year 2001

11292 tritium 20.5% year 2001

18 17.85 17.85 West Vantage 5H no data for year 2001 8973 no data for tritium, year 2001

18 40* * peak linear fuel rating

40* * peak linear fuel rating

West Standard 2612 tritium 99.9% year 2002

4143 tritium 98.6% year 2002

18 17.3 17.3 West XL 11688 tritium 99.9% year 2001

43586 tritium 99.9% year 2003

20838 tritium 5.2 % year 2001

7370 tritium 23.8% year 2003

18 17.3 17.3 West XL 33881 tritium 99.9% year 2001

15385 tritium 99.9% year 2003

6933 tritium 43.0 % year 2001

3963 tritium 27.0% year 2003

18 19.4 19.4 ANF

18 14.5 14.5

18 21.16 21.16 West SIF 18100 tritium 99.9%, year 1994 (1/2 of total for 2 units)

18350 tritium 99.9%, year 1996 (1/2 of total for 2 units)

5400 tritium 6%, year 1994 (1/2 of total for 2 units)


18 21.16 21.16 West SIF 18100 tritium 99.9%, year 1994 (1/2 of total for 2 units)




24 4.31MW/Assembly

4.50MW/Assembly



Siemens 9x9-2 2446 tritium 99.9%, year 2002

5065 tritium 99.9%, year 2002

24 4.31MW/Assembly

4.50MW/Assembly



Siemens 9x9-2

24 18.83 18.83 B&W Mark B8

18 18 18 OFA/LOPAR 13900 tritium 99.9%, year 1994 (1/2 of total for 2 units)

no data 572 tritium 5%, year 1994 (1/2 of total for 2 units)

no data

18 18 18 OFA/LOPAR 13900 tritium 99.9%, year 1994 (1/2 of total for 2 units)

no data 572 tritium 5%, year 1994 (1/2 of total for 2 units)

no data



GE 7x7/ GE 98 P8DWB

18 17.83 17.83 Vantage+/Performance+ 27800 tritium 99.9%, year 1994

21400 tritium 99.9%, year 1996

6120 tritium 18%, year 1994

536 tritium 96%, year 1996

18 17.8 17.8 27400 tritium 99.9%, year 1992 (1/2 of total for 2 units)




18 17.8 17.8 27400 tritium 99.9%, year 1992 (1/2 of total for 2 units)




18 17.52 17.52 CEA 12562 tritium 99.9% year 2001

49362 tritium 99.9% year 2003

4965 tritium 69.0 % year 2001

86593 tritium 3.0% year 2003

18

18 17.88 17.88 V5H no data for year 2000 22237 tritium 99.9%, year 2002


12 43 4.35MW/Assembly

43 4.56MW/Assembly


MWt

USA Wolf Creek WEC 09.03.85 PWR 3411

Slovenia Krsko WEC 01.01.83 PWR 1882

*S- Stretch power (uprates typically up to 7 %, within the design capacity plant). Typically involve changes to instrumenMU-Measurement uncertainty (uprates are less than 2 %e achieved by implementing enhanced techniques for calculaE-Extended power (uprates greater than stretch. Require significant modifications to major BOP equipment) **Second thermal uprate***Third thermal uprate

PWRBWR





18 17.85 17.85 Standard/Vantage 5H 16700 tritium 99.9% year 1992

no data 12040 tritium 5%, year 1992

no data

15-18 17.62 17.62 Vantage5/Vantage+ 10800 tritium 99.9% year 1999

7751 tritium99.9% year 2001




MWt

Belgium Doel 2 WEC 12.01.75 PWR 1192

Doel 3 FRAM 10.11.82 PWR 2785

Tihange 1 ACLF 09.01.75 PWR 2665

Tihange 2/1995 ACLF 06.01.83 PWR 2785

Tihange 2/2001 ACLF 06.01.83 PWR 2785

Finland Loviisa 1 AEE 05.09.77 PWR 1375

Loviisa 2 ASEA 01.05.81 PWR 1375





Germany Brokdorf KWU 01.11.86 PWR 3765

Emsland KWU 06.20.88 PWR 3765

KWU 06.20.88 PWR 3765

Grafenrheinfeld KWU 06.17.82 PWR 3765

Grohnde KWU 02.01.85 PWR 3765

KWU 02.01.85 PWR 3765

Gundremmingen B KWU 07.01.84 BWR 3840

Gundremmingen C KWU 01.01.85 BWR 3840

Isar-1 KWU 03.21.79 BWR 2575

Isar-2/1991 KWU 04.09.88 PWR 3765

Isar-2/1998 KWU 04.09.88 PWR 3765

Neckar-2/1991 KWU 04.15.89 PWR 3765

Neckar-2/2005 KWU 04.15.89 PWR

Philippsburg-2.1991 KWU 12.17.84 PWR 3765



Unterwester KWU 10.01.79 3733

Hungary PAKS1 AEE 11.03.83 PWR 1375

PAKS2 AEE 09.21.84 PWR 1375

PAKS3 AEE 11.03.86 PWR 1375PAKS4 AEE 11.19.87 PWR 1375

Japan Higasidory 05.03.05 BWR ?Shhika HIT 07.30.93 BWR 1593

Korea Kori 3 WEC 10.01.86 PWR 2775

Kori-4 WEC 04.29.86 PWR 2775

YGN 1 WEC 08.25.86 PWR 2775

YGN 2 WEC 06.10.87 PWR 2775

8. Occupational dose in operating year 9. Occupational dose in outage 10. Dose received during uprate8.1 Before the uprate 8.2 After the uprate 9.1 Before the uprate 9.2 After the uprate

manSv manSv manSv manSv man-Sv

0.261 (ISOE) 0.28 0.237 (ISOE) 0.214 (outage)+0.195 (SGR)

1.202 (ISOE 3 year) 0.879 (ISOE, 3 year) 1.03 (ISOE 3 year) 0.75 (ISOE 3 year) 3.30 (1993, ISOE annual)

1.282 (ISOE 3 year) 0.677 (ISOE, 3 year) 1.145 (ISOE 3 year) 0.57 (ISOE 3 year) 3.22 (1995, ISOE annual) 3.09 outage

1.432 (ISOE 3 year) ) 0.986 (ISOE, 3 year) 1.140 (ISOE 3 year) 0.858 (ISOE 3 year) 1.41 (1995, ISOE annual) 1.21 outage

0.488 (ISOE 3 year) 0.752 (ISOE, 2 year) 0.404 (ISOE 3 year) 0.658 (ISOE 2 year) 1.54 (2001, ISOE annual) 1.45 outage

0.979 (ISOE 3 year) 1.095 (ISOE 3 year) 0.901 (ISOE 3 year) 1.042 (ISOE 3 year) 0.869 (1998, ISOE annual) 0.82 outage


0.502 (ISOE 3 year) 0.677 ISOE 3 year) no data no data 0.620 (1984, ISOE annual)


0.410 (ISOE 3 year) 0.677 ISOE 3 year) no data no data 0.620 (1984, ISOE annual)


0.078 (ISOE 2 year) 0.148 (ISOE 3 year) 0.068 (ISOE 2 year) 0.128 (ISOE 3 year) 0.149 ( 1990, ISOE annual) 0.130 outage

0.657 (ISOE 3 year) 0.835 (ISOE 3 year) no data 0.750 (ISOE 2 year) 0.619 (1990, ISOE annual)



0.093 (ISOE 2 year) 0.224 (ISOE 3 year) 0.053 (ISOE 2 year) 0.161(ISOE 3 year) 0.262 (1991, ISOE annual) 0.133 outage

no data no data no data no data no data




1.292 (ISOE 3 year) 1.054 (ISOE 3 year) 1.087 (ISOE 3 year) 0.877 (ISOE 3 year) 1.350 (2000, ISOE annual) 1..094 outage


MWt

Mexco LV 1 GE 07.29.90 BWR 1931

LV 2 GE 04.10.95 BWR 1931

Spain Almaraz-1 WEC 09.01.83 PWR 2696

Almaraz-2 WEC 07.01.84 PWR 2696

Asco-1/2000 WEC 12.01.84 PWR 2696

Asco-1/2003 WEC 12.01.84 PWR 2696

Asco-2/1999 WEC 06.01.85 PWR 2696

Asco-2/2004 WEC 06.01.85 PWR 2696







Sweden Forsmark-1 ASEA 12.01.80 BWR 2711

ASEA 12.01.80 BWR 2711


ASEA 07.01.81 BWR 2711


ASEA 08.01.85 BWR 3020



ASEA 08.15.85 BWR 3020

Ringhals-1 ASEA 01.01.76 BWR 2270

ASEA 01.01.76 BWR 2270



WEC 09.01.81 PWR 2783






0.575 (ISOE 3 year) no data 0.430 (ISOE 3 year) no data 0.454 (2003, ISOE annual) 0.425 outage


0.705 (ISOE 3 year) 0.490 (ISOE 3 year) 0.574 (ISOE 3 year) 0.410 (ISOE 3 year) 0.595 (2000, ISOE annual) 0.605 ??outage

0.457 (ISOE 3 year) 0.494 (only 2004) 0.431 (ISOE 3 year) 0.448 (only 2004) 0.669 (2003, ISOE annual) 0.543 outage 0.102 manSv modifications (mainly to the new access platforms in loop B) and0.071 manSv head vessel replacement. replacement of the SVR (RadiologicalSurveillance System) has been performed.


0.354 (ISOE 3 year) no data 0.309 (ISOE 3 year) 0.716 (2004, ISOE annual) 0.614 outage

0.805 (ISOE 3 year) 3.329 (ISOE 3 year) no dana no data 3.424 (1988, ISOE annual)

1.812 (ISOE 3 year) 2.493 (ISOE 3 year) 1.478 (ISOE 3 year) 1.325 (ISOE 3 year) 0.469 (1998, ISOE annual) 0.032 unplanned outage

1.920 (ISOE 3 year) 3.085 (ISOE only year 2003)


2.154/2.795 (2002)

1.966 (ISOE 3 year) 0.700 (no outage) 1.433(ISOE 3 year) 3.085 (2003, ISOE annual doze) 2.625 outage




0.863 (outage)/0.964 (annual) 2002

0.523 (ISOE 3 year) 0.551 (ISOE 3 year) no data no data 0.501 (1986, ISOE annual)


1.153 (ISOE 3 year) 0.864 (ISOE 3 year) 0.981 (ISOE 3 year) 0.719 (ISOE 3 year) 0.817 (1989, ISOE annual) 0.678outage






MWt

Switzerland Leibstadt/1985 GE 12.15.84 BWR 3012

Leibstadt1998 GE 12.15.84 BWR 3012

Leibstadt.1999 GE 12.15.84 BWR 3012






Gösgen/1985 KWU 11.01.79 PWR 2808

Gösgen/1992 KWU 11.01.79 PWR 2808

USA ANO-2 CE 03.26.80 PWR 2815












Brunswick 1/1996 GE 03.18.77 BWR 2436

Brunswick 1/2002 GE 03.18.77 BWR 2436

Brunswick 2/1996 GE 11.03.75 BWR 2436



0.515 (ISOE only year 1984) 2.329 (ISOE 3 year) no data no data 1.484 (1985, ISOE annual)






1.029 (2002, ISOE annual) 0.492 outage




1.467 (ISOE 3 year) 1.157 (ISOE 3 year) no data 0.988 (ISOE 3 year) 0.957 (1992, ISOE annual) 0.776 outage

year 2002




0.022 (2001, ISOE annual)


0.773 (ISOE 3 year) 0.886 (ISOE 2 year) 0.650 (ISOE 3 year) 0.794 (ISOE 2 year) 0.105 (2001, ISOE annual)



4.911 (ISOE 3 year) 1.293 (ISOE 3 year) 3.920 (ISOE only year 1995)

0.676 (ISOE 3 year) 3.178 (1996, ISOE annual) 2.110 outage


0.622 (ISOE 3 year) no data 2.100 (2002, ISOE annual)



MWt

USA Brunswick 2/2002 GE 11.03.75 BWR 2436

Byron 1 WEC 09.16.85 PWR 3411

Byron 2 WEC 08.21.87 PWR 3411

Callaway WEC 04.09.85 PWR 3411

Calvert Cliffs 1/1977 CE 05.08.75 PWR 2560Calvert Cliffs 1/2006 CE 05.08.75 PWR 2560Calvert Cliffs 2/1977 CE 04.01.77 PWR 2560

Calvert Cliffs 2/2006 CE 04.01.77 PWR 2560

Clinton GE 11.24.87 BWR 2894

Comanche Peak 1 WEC 08.13.90 PWR 3411

Comanche Peak 2 WEC 08.03.93 PWR 3411Comanche Peak 2Crystal River 3 BW 03.13.77 PWR 2544

D.C. Cook 1 WEC 05.07.85 PWR 3250

D.C. Cook 2 WEC 03.13.86 PWR 3411Diablo Canyon 1 WEC 05.07.85 PWR 3338







Fermi 2 GE 01.23.88 BWR 3293

Fitzpatrick GE 07.28.75 BWR 2436

Fort Calhoun/1980 CE 09.26.73 PWR 1420Fort Calhoun/2004 CE 09.26.73 PWR 1420

Fort Calhoun/2006 CE 09.26.73 PWR 1420

Ginna WEC 07.01.70 PWR 1520





0.656 (2002, ISOE annual)





no data 3.303 (ISOE 3 year) no data no data no data (1977)


0.591 (ISOE 3 year) no data 2.081 (2002, ISOE annual) 1.613 outage

0.882 (ISOE 3 year) 0.303 (ISOE 2 year) 0.749 (ISOE 2 year) no outages (2 year) 1.105 (2001, ISOE annual) 1.060 outage

1.013 (ISOE 3 year) 0.704 (ISOE 3 year) 0.732 (ISOE 3 year) 0.551 (ISOE 2 year) 1.086 (1999, ISOE annual) 1.017 outage1.060 (ISOE 3 year) 1.156 (ISOE 2 year) 1.017 (ISOE 2 year) 0.853 (ISOE 2 year) 0.045 (2001, ISOE annual) 1.995 (ISOE 2 year) 1.266 (ISOE only year

2003) 1.735 (ISOE 2 year) 1.210 (ISOE only year

2003) 0.050 (2002, ISOE annual)

no data year 2002

no data year 20031.772 (ISOE 3 year) 0.055 (ISOE 2 year) 1.674 (ISOE 3year) no outages (2 year) 0.904 (2000, ISOE annual) no data for outage





1.745 (ISOE 3 year) 0.979 (ISOE 3 year) 1.270 (ISOE 3year) 0.273 (ISOE 3 year) 2.306 (1998, ISOE annual) 1.958 outage

1.486 (ISOE 3 year) 1.923 (ISOE 3 year) 0.728 (ISOE 3year) 1.464 (ISOE 3year) 2.013 (1998, ISOE annual) 1.830 outage

1.887 (ISOE 3 year) 0.919 (ISOE 2 year) no data no data 2.450 (1992, ISOE annual) no data for outage


2.777 (ISOE 3 year) 3.693 (ISOE 3 year) no data no data 6.680 (1980, ISOE annual) no data for outage1.948 (ISOE 2 year) no data 1.753 (ISOE2 year) no data no data (2004)


MWt

USA Grand Gulf GE 07.01.85 BWR 3833

H. B. Robinson/1979 WEC 03.07.71 PWR 2200H. B. Robinson/2002 WEC 03.07.71 PWR 2200

Hatch 1/1995 GE 01.01.76 BWR 2436

Hatch 1/1998 GE 01.01.76 BWR 2436

Hatch 1/2003 GE 01.01.76 BWR 2436

Hatch 2/1995 GE 09.05.79 BWR 2436

Hatch 2/1998 GE 09.05.79 BWR 2436

Hatch 2/2003 GE 09.05.79 BWR 2436

Hope Creek GE 12.20.86 BWR 3293





Kewanee /2003 WEC 06.16.74 PWR 1650

Kewanee /2004 WEC 06.16.74 PWR 1650





Millstone 2 CE 12.26.75 PWR 2560

Monticello GE 06.30.71 BWR 1670

Nine Mile Point 2 GE 03.11.88 BWR 3323



Palisades CE 12.31.71 PWR 2530

Palo Verde 1/1996 CE 02.13.86 PWR 3800

Palo Verde 1/2005 CE 02.13.86 PWR 3800

Palo Verde 2/1996 CE 09.19.86 PWR 3800

Palo Verde 2/2003 CE 09.19.86 PWR

Palo Verde 3/1996 CE 01.15.88 PWR 3800

Palo Verde 3/2005 CE 01.15.88 PWR 3800





Perry GE 11.18.87 BWR 3579

Pilgrim GE 12.09.72 BWR 1998






1.022 (ISOE 3 year) no outage in year 2003 1.764 (2002, ISOE annual) 1.525 outage

8.777 (ISOE 3 year) 13.370 (ISOE 3 year) no data no data 11.880 (1979, ISOE annual) no data for outageyear 2002

3.478 (ISOE 3 year) 2.964 (ISOE 3 year) no data 2.358 (ISOE 3 year) 1.817(1995, ISOE annual)

3.393 (ISOE 3 year) 1.344 (ISOE 3 year) 2.358 (ISOE 3 year) 1.330 (ISOE only year 2001)

0.530(1998, ISOE annual)

1.391 (ISOE 3 year) no dana 1.022 (ISOE 3 year) no data 0.380 (2003, ISOE annual) 1.525 outage

3.478 (ISOE 3 year) 2.019 (ISOE 3 year) no data 1.413 (ISOE 3 year) 3.080(1995, ISOE annual) 3080 outage ??

2.123 (ISOE 3 year) 1.882(ISOE 3 year) 2.358 (ISOE 3 year) 1.615 (ISOE only year 2001)

2.770(1998, ISOE annual) 2.240 outage



no data year 2003

no data year 2002

no data year 2003

1.933 (ISOE 3 year) 1.998 (ISOE 2 year) 0.693 (ISOE 3 year) 0.369 (ISOE 2 year) 0.585 (2000, ISOE annual) no outage

2.037 (ISOE 3 year) 0.659 (ISOE 2 year) no planned outages only forced the d collective dose

no planned outages only forced the d collective dose


1.479 (ISOE 3 year) 1.338 (ISOE 3 year) no outage in year 1995 0.757 (ISOE 3 year) 1.805 (1996, ISOE annual) 1.380 outage


6.770 (ISOE 3 year) 8.600 (ISOE 3 year) no data no dana 4.270 (1979, ISOE annual) no data for outage





no data for year 2004 no data for year 2004 no data for year 2004 no data for year 2004 no data for year 2004





no data no data no data


0.682 (ISOE 3 year) 1.389 (1996, ISOE annual) no outage

3,313 (ISOE 3 year) 1.336 (ISOE 3 year) no data 0.421 (ISOE 3 year) 2.890 (1994 ISOE annual) no data outage

1.321 (ISOE 3 year) no data 0.433 (ISOE 3 year) no data 2.719 (2002 ISOE annual) 2.107 outage

2.720 (ISOE 3 year) 1.860 (ISOE 3 year) no data 1.034 (ISOE 3 year) 2.995 (1995 ISOE annual) 2.010

1.321 (ISOE 3 year) no data no planned outages no data 0.612 (2002 ISOE annual) no outage

2.133 (ISOE 3 year) 1.639 (ISOE 2 year) 1.656 (ISOE 3 year) 0.970 (ISOE 2 year) 0.558 (2000 ISOE annual) no outage

3.325 (ISOE 3 year) 1.087 (ISOE 2 year) 2.489 (ISOE 3 year) 0.620 (ISOE 2 year) 0.506 (2000 ISOE annual) no outage




MWt

USA Quad Cities 1 GE 02.18.73 BWR 2511

Quad Cities 2 GE 03.10.73 BWR 2511

River Bend/2000 GE 06.16.86 BWR 2894

River Bend/2003 GE 06.16.86 BWR 2894

Salem 1/1986 WEC 06.30.77 PWR 3344

Salem 1/2001 WEC 06.30.77 PWR 3344

Salem 2 WEC 10.13.81 PWR 3411



Seabrook1/2005 WEC 08.19.90 PWR 3411Seabrook1/2006 WEC 08.19.90 PWR 3411Sequoyah 1 WEC 07.01.81 PWR 3411

Sequoyah 2 WEC 06.01.82 PWR 3411

Shearon Harris WEC 05.02.87 PWR 2775



St. Lucie 1 CE 12.21.76 PWR 2560

St. Lucie 2 CE 08.08.83 PWR 2560

Surry 1 WEC 12.02.72 PWR 2441

Surry 2 WEC 05.01.73 PWR 2441





TMI-1 BW 09.02.74 PWR 2535



Vermont Yankee GE 11.29.72 BWR 1593

V. C. Summer WEC 12.02.72 PWR 2775



Waterford 3/2002 CE 09.24.85 PWR 3390

Waterford 3/2005 CE 09.24.85 PWR 3390

Watts Bar WEC 05.27.96 PWR 3411

WNP-2 BWR 3323





0,708 (2001 ISOE annual) no outage



0.663 (2001 ISOE annual) no outage

2.249 (ISOE 3 year) 1.214 (ISOE 2 year) 1.387 (ISOE 3 year) 0.756 (ISOE 2 year) 2.016 (2000 ISOE annual) 1.329 outage

no data no data no data no data no all data for 2003






0367 (ISOE 3 year) 1.350 (ISOE only year 2002)

0.068 (2001, ISOE annual) no outage


no data no outage in year 2002 0.578 (2001, ISOE annual) no outage



0.069 (2001, ISOE annual) no outage

no all data year 2002 no all data year 2002 no all data year 2002 no all data year 2002 no all data for year 2002

no all data for year 2002 no all data for year 2002 no all data for year 2002 no all data for year 2002 no all data for year 2002

0.745 (ISOE 2year) 0.067 (ISOE only year 2002)

0.588 (ISOE 3 year) no outage in year 2001) 2.522 (2001, ISOE annual) 2.412 outage




6.315 (ISOE only year 1984) 3.422 (ISOE 3 year) no data no data 6.720 (1985, ISOE annual) no data for outage

2.173 (ISOE 3 year) 1.461 (ISOE 3 year) no data 0.864 (ISOE 3 year) 2.227 (1995, ISOE annual) no outage


2.501 (ISOE 3 year) 1.401 (ISOE 3 year) no data 0.578 (ISOE 3 year) 2.665 (1995 ISOE annual) 1.830 outage




2.610 (ISOE 3 year) 1.860 (ISOE 3 year) no data 0.578 (ISOE 3 year) 2.210 (1994 ISOE annual) no data for outage


0.467 (ISOE 3 year) no outage in year 2002 2.390 (2001 ISOE annual) 1.890 outage


1.738 (ISOE 3 year) 1.037 (ISOE 3 year) no data 0.927 (ISOE 3 year) 0.137 (1996, ISOE annual) no outage


2.110 (ISOE 3 year) 1.069 (ISOE 3 year) no outage in year 1995) 0.954 (ISOE 3 year) 1.060 (1996, ISOE annual) 0.890 outage

2.090 (ISOE 3 year) 1.236 (ISOE 3 year) no data 1.095 (ISOE years 95 and 96)

1.833 (1993, ISOE annual) no data for outage

2.090 (ISOE 3 year) 1.657 (ISOE 3 year) 1.947 (ISOE years 95 and 96)

no data 1.800 (ISOE years 95 and 96)

1.833 (1993, ISOE annual) no data for outage

no all data for year 2002 no all data for year 2002 no all data for year 2002 no all data for year 2002 no data year 2002

no data year 2005





2.539 (ISOE 3 year) 4.273 (1995 ISOE annual) 2.973 outage


MWt

USA Wolf Creek WEC 09.03.85 PWR 3411

Slovenia Krsko WEC 01.01.83 PWR 1882

*S- Stretch power (uprates typically up to 7 %, within the design capacity plant). Typically involve changes to instrumenMU-Measurement uncertainty (uprates are less than 2 %e achieved by implementing enhanced techniques for calculaE-Extended power (uprates greater than stretch. Require significant modifications to major BOP equipment) **Second thermal uprate***Third thermal uprate

PWRBWR





2007:01 Statens ansvar för slutförvaring av använt kärnbränsle

SKI och SSI

2007:02 Strålmiljön i SverigeAvdelningen för beredskap och miljöövervakning Pål Andersson et.al. 310 SEK

2007:03 personalstrålskydd inom kärnkraft- industrin under 2005

Avdelningen för personal- och patientstrålskydd Stig Erixon, Karin Fritioff, Peter Hofvander, Ingemar Lund, Lars Malmqvist, Ingela Thimgren och Hanna Ölander Gür 70 SEK

2007:04 recent research on emf and Health risks. fourth annual report from SSI’s Independent expert group on electromagnetic fields, 2006

Avdelningen för beredskap och miljöövervakning 110 SEK

2007:05 doskatalogen för nukleärmedicinprojekt; SSI p 1426.04

Avdelningen för personal- och patientstrålskydd Sigrid Leide-Svegborn, Sören Mattsson, Lennart Johansson, Per Fernlund och Bertil Nosslin 90 SEK

2007:06 personalstråldoser inom vård, forskning och icke kärnteknisk industri i Sverige under 1999-2005

Avdelningen för personal- och patientstrålskydd Catarina Danestig Sjögren 100 SEK

2007:07 Inquiry into the radiological consequences of power uprates at light-water rectors worldwide

Avdelningen för personal- och patientstrålskydd Tea Bilic Zabric, Bojan Tomic, Klas Lundgren and Mats Sjöberg 290 SEK

SSI-rapporter 2007pporter 2007r 2007 SSI reports 2007

Adress: Statens strålskyddsinstitut; S-171 16 StockholmBesöksadress: Solna strandväg 96Telefon: 08-729 71 00, Fax: 08-729 71 08

Address: Swedish Radiation Protection AuthoritySE-171 16 Stockholm; SwedenVisiting address: Solna strandväg 96Telephone: + 46 8-729 71 00, Fax: + 46 8-729 71 08

www.ssi.se

S TATENS STRÅLSKYDDSINSTITUT, SSI, är en central tillsyns- myndighet som verkar för ett gott strålskydd för människan och miljön, nu och i framtiden.

SSI sätter gränser för stråldoser till allmänheten och för dem som arbetar med strålning, utfärdar föreskrifter och kontrollerar att de efterlevs. SSI håller beredskap dygnet runt mot olyckor med strålning. Myndigheten informerar, utbildar och utfärdar råd och rekom- mendationer samt stöder och utvärderar forskning. SSI bedriver även internationellt utvecklingssamarbete.

Myndigheten, som sorterar under Miljödepartementet, har 110 anställda och är belägen i Solna.

THE SWEDISH RADIATION PROTECTION AUTHORITY (SSI) is a central regulatory authority charged with promoting effective radiation protection for people and the environment today and in the future.

SSI sets limits on radiation doses to the public and to those that work with radiation. SSI has staff on standby round the clock to respond to radiation accidents. Other roles include information, education, issuing advice and recommendations, and funding and evaluating research.

SSI is also involved in international development cooperation. SSI, with 110 employees located at Solna near Stockholm, reports to the Ministry of Environment.

Inquiry into the radiological consequences of power uprates at light-water rectors worldwide

Documents

Transcript of Inquiry into the radiological consequences of power uprates at light-water rectors worldwide